Simple and stable waste streams

Fuel cycle operations can be performed using a variety of fabrication processes.

Enrichment can be carried out by centrifugation or by gaseous diffusion, and fuel fabrication may involve different methods or different fissile materials (e.g. MOX fuel).

It would be unrealistic to specify recommendations applicable in all circumstances. Two examples are cited to illustrate what could be an acceptable methodological approach for declaring the radionuclides found in the waste packages. Each participant in the process (waste producer, regulatory authorities, and organization responsible for ultimate disposal) is responsible for adapting these examples to other comparable situations.

4.2.1.1. Natural uranium enrichment by gaseous diffusion

The gaseous diffusion process for enrichment of natural uranium uses uranium hexafluoride UF6. Uranium is a mixture of three natural isotopes, whose molar concentrations—and therefore whose relative radioactivities—vary throughout the diffusion cascade.

In contact with water vapour, even in trace amounts, UF6 hydrolyses to form aerosols that are deposited on surfaces encountered by the fluid, resulting in radioactive surface contamination. All tubes and vessels in contact with the process fluid are contaminated to some extent.

Maintenance and servicing operations in the enrichment plant or facilities used for disassembly, cleaning, adjustment, repair, replacement, etc., thus produce contaminated waste and residues that must be collected and eliminated.

A distinction is made between compactable and non compactable waste, and between wet and dry waste, by source segregation.

Production is continuous, although a few interruptions occur for some types of plant components (e.g. valves, pipes, rotating machinery, etc.) during preventive maintenance activities.

General waste characteristics

Compactable waste mainly consists of materials used for the following:

⎯ in situ decontamination of parts after disassembly or maintenance of process tubes and equipment; these materials are mainly rags, cotton swabs, synthetic absorbent fabrics;

⎯ protection of surfaces against contamination during maintenance operations: PVC or polyethylene sheeting and bags;

⎯ protection of personnel against contamination: gloves, clothing, PVC or polyethylene;

⎯ treatment (soaking) of filter canisters and permanent or mobile purification devices; the latter include a combination of filter, rubber gasket, PVC, etc.

A mean physical composition can be calculated for the compactable waste (percentage of PVC, polyethylene, latex, paper, cotton, fabrics, etc).

Non compactable waste consists mainly of scrapped parts and materials:

⎯ metal, rubber or synthetic gaskets,

⎯ complete or cut-up metal parts that would be significantly more expensive to decontaminate than to ship to a disposal site,

⎯ absorbent materials,

⎯ rubble,

⎯ soil.

The traceability of the non compactable waste form placed in each drum must be ensured.

Nature and activity of the radionuclides

Contamination in the waste consists primarily of aerosols or fine particles deposited on the previously mentioned physical surfaces after contact with the atmosphere or other surfaces. It can also be due to impregnation after contact with a liquid. The contamination is thus generally not strongly bound to the surface. The contaminating agent is uranium:

⎯ mainly as uranium oxyfluoride (UO2F2) resulting from hydrolysis of UF6; the oxyfluoride may be anhydrous but is generally found as (UO2F2,xH2O) or even (UO2F2,xH2O,yHF);

⎯ the form of complex fluorides produced by the reaction of UF6 with fluids such as oil or with gaseous chlorofluoride compounds.

⎯

While oxyfluorides are highly soluble in water, acidic and basic solutions, the products of the second category are very often insoluble.

Methodology for declaring the radionuclides in waste materials

Uranium is a mixture of three natural isotopes (234, 235 and 238); the ²³⁵U concentration varies across the diffusion cascade.

A mean uranium-235 concentration can be determined based on the experience acquired over several years of plant operation. The declaration must be thoroughly prepared on the basis of representative waste samples from all the process facilities taken at frequent intervals throughout the year. In this context, the results of sample measurements of gaseous discharges from the plant to the atmosphere would appear to be suitable indicators.

Having evaluated the mean ²³⁵U concentration, the corresponding ²³⁴U concentration must then be calculated; the remainder represents the ²³⁸U concentration.

The mean molar isotopic composition of the representative waste mixture is used to calculate the specific activity of the mixture and to declare the radioelements accompanying uranium. Uranium is always found together with its short half-life decay products, ²³⁴Th and

234Pa, the first daughter isotopes of ²³⁸U.

These radioisotopes can be present in the waste:

⎯ either in nominal proportions with respect to uranium—i.e., at radioactive equilibrium—when the contamination is due exclusively to uranium,

⎯ or in slight excess, when the contamination arises mainly from nonvolatile Th and Pa volatile fluorides that generally deposit at the bottom of process vessels or at singularities in the flow lines.

In the second case, once they are isolated, the activity of the fluorides decreases at the half-life of ²³⁴Th (24 days); i.e., between the moment the waste is produced and the moment it reaches the disposal site, the βγ activity is no longer significant.

The radioactivity of the waste package is calculated:

⎯ from γ spectrometry measurements (on the principal ²³⁵U line) for each package shipped,

— from a mathematical transfer function allowing for the influence of the parameters likely to affect the measurement (weight, type and density of the waste, measurement distance),

— from the calculated standard spectrum (²³⁴Th, ²³⁴Pa, ²³⁸U).

In a plant fabricating MOX fuel assemblies for nuclear power stations, production is organized in fabrication campaigns to supply the client with fuel assemblies meeting the nuclear specification requirements (Pu and U isotopic composition and content at a given reference date).

The finished product is a mixture of uranium and plutonium with well-defined characteristics, making it a simple matter to calculate the standard spectrum for each production campaign. The reference spectrum is thus the standard spectrum for the production

c

ampaign in progress at the moment the waste containers are filled. This data must be conserved for traceability purposes.

If the waste arises from a zone in which only one of the two constituents is handled (e.g.

t

he uranium zone), it is advisable to specify the standard spectrum for that zone and declare the radionuclides in the waste accordingly.

The standard reference spectrum is validated on the following assumption If the process used to produce fuel rods with the contractual nuclear characteristics is fully controlled, there is a strong probability that the waste produced will have identical characteristics.

Consequently, in routine production the waste drums are measured, and the radionuclides declared are those identified in the standard reference spectrum. Conversely, additional measurements are required in the event of a malfunction liable to affect the standard reference spectrum.

In this type of production the beta-emitting radionuclides spectrum is often simple (e.g.

99Tc) and can be determined with respect to the total alpha activity rather than with respect to a single key nuclide.

Standard spectrum management

When the alpha activity (at the reference date) of the MOX fuel varies minimally from one production campaign to the next (this criterion must be the subject of a common agreement between the waste producer, the regulatory authority and/or the organization responsible for final disposal), the standard reference spectrum of the preceding campaign can be reused.

This procedure allows successive production campaigns to be organized at optimum cost without significantly affecting the quality of the data characterizing the waste.

4.2.2. Nuclear power plant

This waste results from treatment of cooling water, equipment decontamination, and routine facility maintenance. As discussed in Section 3, part of the traceable waste streams from NPPs can be classified as simple and stable. The main characteristics of this type of waste are:

4.2.1.2. MOX fuel fabrication General methodology

— sampling is easy or not necessary

— easy to measure using NDA or DA methods

— simple spectrum

— good relation between dose rate and activity (scaling factor method is applicable)

— physical and chemical properties are known or easy to measure (DA)

Typical examples of this category are filters, neutron sources, calibration sources, and defectoscopy sources.

An ISO standard guide “Scaling factor to determine the radioactivity of low and intermediate radioactive waste packages generated at NPP” [18] is in progress. The contents of this guide describe the principles of the scaling factor methodology applied to NPP, the sampling requirements (representation, rejection of outliers, records of samples) and the evaluation methodology for scaling factors (evaluation by linear or non-linear relationship;

selection of key-nuclides; integration of corrosion products, fission products and alpha emitting nuclides, and accuracy). This scaling factor methodology for NPP operational waste is suitable when many samples can be taken from the same waste stream. This methodology is not be applicable to current reprocessing waste, because this typically involves many different campaigns.

4.2.3. Spent fuel

Spent fuel actually consists of two barriers that may prevent release of radionuclides: the cladding and the fuel matrix itself. Some programs take credit for the protection cladding may offer. However, other programs do not consider cladding as a barrier mostly because the final state of the cladding after irradiation is not always well known. For example, cladding failures in the fuel can usually be identified by radioactivity release to the reactor coolant during cool down and depressurization of the reactor. While this will identify the presence of failed fuel, additional techniques such as sipping are sometimes needed to accurately identify the failed assembly. Still, it might not be known how many rods within that assembly have failed. The presence of tramp material in the core may also make it difficult to detect pin-hole defects in cladding. A combination of sipping, visual inspection, eddy current or ultrasonic testing may be performed to identify the percentage of failed fuel rods.

Even if the number of failed rods can be accurately determined and documented, it is difficult to know the state of intact fuel rods. During reactor operation, the cladding undergoes oxidation and hydriding that can affect their performance. As the fuel is pushed to higher burnups, these reactions become more significant. Thus, it is not clear what the remaining wall thickness is for the intact rods. The uncertainty in how many additional rods might fail during handling, transportation, or storage due to these incipient failures is the main reason many programs do not take credit for cladding as a barrier to radionuclide release.

Once the cladding is breached, fission gases and other volatiles in the fuel/clad gap and plenum may be released. The fraction of inventory available for release is a function of fuel burn-up and the power (temperature) history. This can be readily calculated using records from the utility and codes such as ORIGEN. Additional radionuclides can then be released if the fuel is exposed to oxidizing conditions either by air or water or even as result of radiolysis under otherwise reducing or anoxic conditions.

4.3. COMPLEX AND STABLE WASTE STREAMS