B.-R. Martens
Bundesamt für Strahlenschutz
P. O. Box 10 01 49, D-38210, Salz gitter

W. Kunz
GNS mbH
P. O. Box 10 12 53, D-45012 Esse

F. W. Ledebrink
Siemens AG
P. O. Box 11 00 60, D-63434 Hanau

Disposal of radioactive waste from the operation and decommissioning of nuclear power plants, medicine, industrial applications and research has become an everyday occurence in Germany since the reopening of the repository at Morsleben in 1994 - despite of the ongoing controversial discussion on the future of nuclear energy. The waste acceptance requirements of this site impose very low limitations to the content of long-lived radionuclides in the waste packages - in particular for alpha emitters. It is not possible therefore to dispose of in this site wastes like waste from the fabrication of mixed-oxide-fuel elements or waste from the reprocessing of spent fuel elements. According to the German plannings these waste types will be disposed of in the Konrad site after the year 2002 - except heat generating high level wastes like vitrified fission products from reprocessing which will be disposed of not before 2012 in the Gorleben site. No waste acceptance requirements have been defined for the Gorleben site; they have been defined but no license has been granted for the Konrad site. On the other hand the mixed-oxide-fuel fabrication plant (MOX plant) of the Siemens AG (Germany) is now in the stage of decommissioning and fission products from German nuclear power plants are vitrified by COGEMA (France) for several years now and by BNFL (UK) starting in October 1997. To avoid unnecessary large-scale non-destructive and/or destructive testing of plutonium-contaminated waste or fission products after conditioning BfS agreed with the waste producers to qualify these conditioning processes for disposal. The prerequisites, procedures and difficulties to qualify these processes are explained.

The main objective of quality assurance (QA) for radioactive waste packages is to provide adequate confidence of fulfilling the waste acceptance requirements of the appertaining treatment, storage or disposal facilities. Within the licensing procedure for the Konrad repository the qualification of conditioning processes has early been identified to be an adequate QA - measure to demonstrate the fulfilment of these requirements and to avoid unnecessary large-scale measurements on waste packages. This qualification corresponds largely to the application of process control quality plans [1] assuring that all waste packages comply with the requirements, that all necessary records are generated, maintained and available, and appropriate levels of control are applied. For the ongoing disposal of short-lived L/ILW in the Morsleben site about 600 conditioning campaigns have been qualified and about 17,000 m³ of radioactive waste have been disposed of there from 1994 to 1997. Similar qualifications of waste conditioning processes for the Konrad and Gorleben sites pose problems as follows:

• Waste acceptance requirements (WAR) for the planned Konrad site have been published first in 1985 and have been revised several times until 1995 by BfS. Any L/ILW containing higher amounts of alpha-emitters should have been conditioned according to these require ments. Within the licensing procedure for this site the environment ministry of Lower Saxony now calls the limitations for fissile materials into question and intends to prescribe lower limits.
• Up to now no waste acceptance requirements have been defined for the Gorleben site. As a first step it was necessary therefore to define quality requirements which seem to or may be relevant for disposal according to present knowledge. Respective values for these properties will not be defined before 2005 as a result of the suitability investigations for this site.

The qualification of the conditioning process for decommissioning wastes from the MOX-fuel fabrication will be taken as an example to explain the difficulties arising for wastes containing higher amounts of alpha-emitters. This decommissioning represents some challenges from radiation protection and economic reasons.

The preliminary waste acceptance requirements of the Konrad site for L/ILW with neglectable heat generation have been discussed elsewere [2]. Concerning the fissile material content the appertaining requirements have been derived from safety analyses for criticality incidents in the operational and post-operational phase. Limitations as follows have been derived (simplified):

• The concentration of fissile materials within the waste product is limited to 50 g/100 l.
• The fissile content of any package is limited to 45 % of its critical mass (H₂O taken as moderating and reflecting material) and the total fissile content of all packages stacked in a row in an emplacement chamber is limited to the critical mass (brines taken as moderating and the host rock as reflecting material). For small packagings like cylindrical concrete or cast iron containers of which up to 39 may be stacked in a row the second requirement results in further limitations.
• Combustible wastes containing more than 1 g of fissile materials per package must be fixed in an incombustible matrix or must be packed in a thermally insulating concrete or tight cast iron container.

There is an ongoing discussion concerning the limitations for the content and concentration of fissile materials within a package: A cementation of fissile materials has been requested if a waste package contains more than 15 g of fissile material and the ability to guarantee an homogeneous distribution of fissile materials within the a. m. limitations is called into question.

The plutonium contaminated areas of the former MOX-fuel plant of the Siemens Company at Hanau (Germany) consist mainly of glove-boxes made from austenitic steel. These boxes may be linked to other boxes for the transfer of MOX-fuel and may contain windows made of glass or plastics, openings for gloves, filters, and for maintenance or take out of materials. They may have an overall volume of several cubic meters. With the objective to reduce personnel exposure and the total project costs the best way to condition this waste for disposal seemed not to compact these boxes or to cut them into shreds but to pack the boxes into containers without destroying them. It is intended to obtain the mechanical and thermal stability necessary for disposal by foaming the boxes from inside with polyurethane and by embedding the boxes with cement into containers [5].

To demonstrate the compliance to the upper limits for the total content and concentration of fissile materials neutron dose rate measurements are planned using a passive neutron coincidence counting system. The detectors comprise up to 30 polyethylene moderated ³He detector modules combined with charge amplifiers and coincidence counting electronics. All measurements are performed from outside of the the glove boxes except for in-situ calibration of the system via the "add-a-source" technique using a Cf-252 source of known activity. The detection limit is below 1 g Pu-240 equivalent if grade is known.

Requirements for containers concern the dimensions, handling, stackability, tightness and stability in case of mechanical and thermal incidents. The demonstration of compliance for the incidents will be discussed in detail.

In case of mechanical incidents the packaging must withstand a drop from a height of 5 m on to a concrete plate which simulates the host rock of an emplacement chamber. Calculations have been performed with the aid of the DYNA3D nonlinear three-dimensional finite element code for solid and structural mechanics [3] giving the following results:

• Mechanical strains of about 25 % at the edges and more at the corners of the outer packaging will damage this carbon steel container.
• The concrete layer surrounding the glove-box will be damaged in part and only deformations will occur in the corner region of the glove-box.

In case of thermal incidents the package must withstand an one hour lasting fire and the temperature raise of the waste product must be limited by thermal insulation to avoid melting or inadvertant release of volatile radionuclides from the waste product. Calculations have been performed in this case with the aid of the finite difference code HEATING 7.2b giving the following results:

• Temperatures at the outer surface of the steel container will reach approximately the temperature of the fire (800 °C) within 10 minutes.
• The temperatures at the outer surface of the glove-boxes will raise a few centigrade after one hour and no melting or inadvertant release will occur therefore in this region.

The formal procedure for the examination of the package design [4] is still outstanding.

Another challenge has been the qualification of the vitrification processes for high level liquid waste from reprocessing of spent fuel. It is planned to dispose of this type of waste in the Gorleben site but up to now no site specific safety analyses have been performed for this salt dome and no waste acceptance requirements have been defined for this repository. BfS and the waste producers agreed therefore to qualify this process based on properties which seem to be or may be relevant for disposal in deep geologic formations according to present knowledge.

For the definition of these properties safety during the operational and post-operational phase of below-ground repositories has been considered as far as this safety depends on the waste. They may be coarsely subdivided into properties of the waste package and waste properties controlled as process parameters during waste conditioning. The vitrification process of COGEMA (France) has been qualified by BfS on the following basis (simplified):

• Total and nuclide specific activities are determined by measurement of samples from the feed tank, dose rate measurements and burn-up calculations
• Fissile materials concentration and the content of neutron absorbers are controlled by adeqaute mixing, neutron measurements and specified composition of the glass product for criticality safety
• Thermal properties like the power and the central temperature of the glass blocks are controlled by calculation via activity taking heat removal in the interim stores and thermal conductivity of the glass product into account
• Dose rate and surface contamination are controlled by measurement, calculation or visual inspection of the packages
• Quality of fixation material is controlled by the glass deliverer and COGEMA in addition for the chemical composition within defined limits
• Composition of raw waste is controlled by analytic determination of its chemical composition and content of annealing residues or oxide content
• Mass fractions waste/frit/additives are controlled within process tolerance by alarm and intervention limits for adequate dosing
• Adequate mixing defined by vitrification/inclusion and homogeneous distribution of the raw waste are controlled by the furnace temperature, the mean refining time and the outlet velocity of the furnace
• Product conditions defined as mass density, transformation temperature and storage below transformation temperature are controlled by the last four properties given before and by adeqate storage conditions
• Hydrolytic stability of the glass product determined by its corrosion behavior in aqueous solutions, its surface area and fraction of soluble phase formations is controlled by the cooling rate of the glass canisters and controlled dosing mentioned above
• Quality of the canister defined by material specifications, weight, marking, dimension, tightness and corrosion resistance is controlled by examination of package design, documentation checking, leackage test, parameters of the lid welding and visual checking
• Package mass is controlled by weighing
• Stackability and handling defined by the number of canisters stackable without deformation, suitable attachment means and response to deformation is controlled by examination of package design (see above), calculation of the degree of filling and determination of the surface enlargement
• Marking of canisters defined by their unique, indelible signature is controlled visually

This qualification should allow for assessment of future waste acceptance requirements without non-destructive and/or destructive testing of the glass canisters. The qualification of BNFL's vitrification plant at Sellafield (UK) has nearly been finished and several audits/inspections have been performed to the work of BNFL and Lloyds Register.

Control measures as described in the previous section are performed by COGEMA. On behalf of the waste producers - including the German utilities - experts of the French organisation Bureau Veritas check the work of COGEMA having unrestricted access to the facilities and documentations generated in the course of waste processing. German experts check the work of COGEMA and Bureau Veritas twice a year by auditing on behalf of BfS. Beside this other German experts check the measuring of dose rates, the packaging of the glass canisters into the transport casks and the closure of the lids on behalf of the controlling authority of the intermediate storage facility at Gorleben giving further assurance for the quality of the redelivered vitrified waste.

1. INTERNATIONAL ATOMIC ENERGY AGENCY, Quality Assurance for Radioactive Waste Packages, Draft IAEA Report, IAEA, Vienna, June 1994
2. P. Brennecke, E. Warnecke, The Konrad waste acceptance requirements: Guidance for radioactive waste conditioning, Proceedings of the Symposium on Waste Management, Tucson, 28 February - 4 March 1993, Vol. 1, Tucson 1993, p. 411 - 415
3. R. G. Whirley, B. E. Engelmann, DYNA3D: A Nonlinear, Explicit, Three-Dimensiomnal Finite Element Code for Solid and Structural Mechanics - User Manual, UCRL-MA-107254 Rev. 1, Lawrence Livermore National Laboratory, November 1993
4. B.-R. Martens, B. Droste, K. E. Wieser: Tests for Packagings for the Disposal of Radioactive Wastes, in: Requirements for waste acceptance and Quality Control, Proceedings of the 2nd International Seminar on Radioactive Waste Products, 28 May - 1 June 1990, BfS-Schriften 1/90, Salzgitter (1990)
5. F. W. Ledebrink, W. Lehr, L. Lasberg: "In-Situ Conditioning of Plutonium-Contaminated Glove-Boxes, 3rd Symposium "Conditioning of Radioactive Operational and Decommissioning Waste", Kontec 1997, A-2-1, March 1997, Hamburg, p. 536-541

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