Long term performance of metal seals for transport and storage casks

Dual purpose casks for the transportation and storage of spent nuclear fuel and other radioactive materials require very high leak tightness of lid closure systems under accident conditions as well as in the long term to prevent activity release. For that purpose metal seals of specific types with an inner helical spring and outer metal liners are widely used and have shown their excellent performance if certain quality assurance requirements for fabrication and assembling are satisfied. Well defined surface roughness, clean and dry inert conditions are therefore essential. No seal failure in a loaded cask happened under these conditions until today. Nevertheless, the considered and licensed operation period is limited and all safety assessments have been performed and approved for this period of time which is 40 years in Germany so far. However, in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long term behaviour with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminium) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behaviour for additional test parameters are explained.     

Leakage of Transport Container Seals During Slow Thermal Cycling to −40°C

Abstract

Measurements of gas leakage rates have been made on Viton and silicone O-ring seals during slow thermal cycling between +10°C and ?40°C. Additional measurements of the sealing force exerted by the seal on the flange have also been made during the tests. Measured leak rates through Viton seals increase to 1 kPa.cm³.s^?1 10^?2 bar.cm³.s^?1 at ?40°C but it is shown that this would not result in loss of particulates through the seal. On increasing the temperature to +10°C, the leak rate decreases to its original value of 5 Pa.cm³.s^?1 (5×10^?5 bar.cm³.s^?1.) Leak rates through silicone seals do not change significantly in the temperature range tested.     

Thermomechanical Finite-element Analyses Of Bolted Cask Lid Structures

Abstract

Analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak tightness in package design assessment according to the transport regulations or in aircraft crash scenarios. In this context BAM is developing methods based on finite elements (FE) to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. In the case of fire it might not be enough to perform only a thermal heat transfer analysis. A complex cask design together with a severe hypothetical time–temperature curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and their counterparts that can be analysed by a so-called thermomechanical calculation. Although it is currently not possible to directly correlate leakage rates with results from deformation analyses, an appropriate finite element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field, respectively. Except for the lid bolts, the geometry of the cask and the mechanical loading is axisymmetric which simplifies the analysis considerably, and a two-dimensional finite-element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modellings on the relative displacements at the seating of the seals. The influence of bolt modelling, thermal properties and the detailed geometry of the two-dimensional finite-element models on the results are discussed.     

Experimental Testing on a Simulated Flask Lid Seal System to Confirm Containment Measuring Procedures

Abstract

British Nuclear Fuels Ltd (BNFL) operates a large number of package types for the transport of irradiated nuclear fuels from customers' utilities to the Sellafield reprocessing facility in Cumbria. All fuel loading operations are carried out under water and consequently package lid sealing systems are saturated. All BNFL package types employ a double seal system on the lid which must be tested before despatch, to confirm containment integrity. The normal test procedure involves drying the interspace between the seals with compressed air before applying a gas pressure and measuring the pressure drop; the reliability of this procedues depends upon the seals being dry. In order to demonstrate the reliability of BNFL's containment testing methods, and to develop operational procedures that ensure acceptable dryness is achieved, an experimental test rig was designed and manufactured. Closely based on a typical package lid seal arrangemens, the test rig allowed leakage paths to be introduced by fine wires fitted across the seal faces. BNFL conducted a series of tests to investigate how the measured leak rate was influenced by the presence of water. Existing drying procedures were evaluated, and shown not to be fully effective in removing all moisture. New drying procedures were subsequently developed, which are totally efficient in drying the test inter-space and ensure that accurate containment measurements can be undertaken.     

Temperature tests on package lid seal materials over one year duration

Abstract

The design of lid seal system of any package transporting irradiated nuclear fuel is a major aspect of the containment safety case under normal and accident conditions. Consequently, when BNFL, now known as British Nuclear Group Sellafield, BN-GS, decided on a change in lid seal material, year long proving trials on the new material were conducted, these simulating actual service and accident temperature conditions. Several years ago, BN-GS with the Rubber and Plastic Research Association (RAPRA) developed an elastomer called EPDM 30H which could perform at temperatures below ?40°C and had excellent resistance to radiation, as later proven by long term testing. EPDM 30H material has been used in many site applications, with its first transport package application being on the Excellox 6 type, commissioned in 1991. Towards the end of the 1990s, BN-GS became aware that the fluorocarbon grade of lid seal material, used on several other packages, may be discontinued by the manufacturers and so they took the decision to replace the seals on the NTL 11 type packages with EPDM 30H. Consequently, BN-GS embarked on an extensive programme of seal testing which went beyond any it had previously carried out. Comprehensive data were available on radiation resistance and performance at low temperatures, but additional data were needed on its behaviour at elevated temperature over periods of about one year. A number of test sets were assembled comprising seals in representative lid seal grooves. Most of the test sets were fitted with EPDM 30H seals but another grade of EPDM seal was also included. Two test sets were continuously maintained at temperatures of 120 and 150°C, for one year. Other test sets were maintained at 60, 90, 120, 140, 160 and 180°C, but periodic inspection and compression set measurements were taken. At the end of one year the continuously heated seals were subjected to a thermal cycle corresponding to the thermal accident safety cases studies. These tests demonstrated that the EPDM 30H material had lower compression set characteristics than the other grade of EPDM and that EPDM 30H was a suitable seal material for all irradiated LWR fuel transport packages currently operated BN-GS.     

Leakage behavior of inflatable seals subject to severe accident conditions

The paper describes tests to determine the leakage behavior of inflatable seals when subjected to containment pressures that exceed the design basis.² Inflatable seals are used to prevent leakage around personnel and escape lock doors in about 10% of the commercial nuclear power plant containment structures in the United States. All of the installations are in either Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) Mark-Ill type containments. This work is a part of an overall effort at Sandia National Laboratories to develop proven techniques for evaluating the performance of Light Water Reactor (LWR) containment buildings for beyond design basis loadings.Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Parameters that were monitored and recorded during each test were the internal seal pressure and temperature, chamber (containment) pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. An empirically based, analytical method is presented to predict the containment pressure at which significant leakage past inflatable seals can be expected.     

Testing of Elastomer Seals using Small-Size Rigs

Abstract

This paper looks at the use of small size seal leakage test rigs to demonstrate the compliance of full size container seals against the IAEA Transport Regulation's limits for activity release for normal transport and accident conditions. The detailed requirements of the regulations are discussed and it is concluded that an appropriate test programme to meet these requirements, using only small size test rigs, can normally be set up and carried out on a relatively short time scale. It is important that any small test rigs should be designed to represent the relevant features of the seal arrangement and the overall test programme should cover all of the conditions, specified by the regulations, for the type, classification and contents of the container under consideration. The parameters of elastomer O-rings, which affect their sealing ability, are considered and those which are amenable to small scale testing or have to be modelled at full size are identified. Generally, the seals used in leakage tests have to be modelled with a full size cross-section but can have a reduced peripheral length.     

The Transport of Aerosols through Ultrafine Leak Paths

Abstract

The aim of the programme has been to provide generic information to enable a valid assessment to be made of the potential leakage of airborne particles through the seals of containers used to transport nuclear materials in which material in powdered form may be present. A series of experiments has been undertaken using model leak paths based on ultrafine capillaries to quantify the penetration behaviour of a variety of aerosols, varying from spherical calibration standards to irregularly shaped particles more typical of those likely to be encountered under transport conditions. These experiments have been supported by the development of a model that explains the findings and enables aerosol penetration behaviour to be predicted over a much wider range of conditions. These studies lend support to the argument that the current IAEA recommended air leakage rate limit of 10–6 Pa·m³·s^?1 is sufficient to prevent the migration of significant particulate-borne radioactivity from a single leak path, although further information is needed about the nature of likely aerosol sources associated with transport movements before a more definitive statement can be made.     

Packaging and security of radioactive material

Abstract

Elastomers are widely used as the main sealing materials for containers for low and intermediate level radioactive waste and as an additional component to metal seals in spent fuel and high active waste containers. The safe encapsulation of the radioactive container inventory has to be guaranteed according to regulation and appropriate guidelines for long term storage periods as well as for temperatures as low as ?40°C during transport. Therefore, the understanding of failure mechanisms that lead to leakage at low temperatures is of high importance. It is known that the material properties of elastomers are strongly temperature dependent. At low temperatures, this is caused by the rubber–glass transition (abbreviated: glass transition). During continuous cooling, the material changes from a rubber-like entropy elastic to a stiff energy elastic behaviour, which allows nearly no strain or retraction. Hence, rubbers are normally used above their glass transition, but the minimum working temperature limit is not defined precisely; this can cause problems during the above noted applications. Therefore, the lower operation temperature limit of elastomer seals must be determined in dependence of the material properties. Differential scanning calorimetry (DSC) and dynamic mechanical analysis (DMA) are combined with standardised measurements to determine the compression set according to ISO 815. To reduce the test time of the standard tests, a faster technique than normally used was developed. Additionally, the breakdown temperature of the sealing function of complete O ring seals was measured in a component test set-up to compare it with the results of the other tests. The experimental set-up is capable of measuring the leakage rate at low temperatures by the pressure rise method. A fluorocarbon rubber (FKM) was selected for this investigation as it is often used for radioactive waste containers. The materials (seals and test sheets) were purchased from a commercial seal producer.     

Radioactive Material Package Seal Tests

Abstract

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the United States (US) Code of Federal Regulations Part 71. The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high and low temperature environments under dynamic and static conditions. A seal technology programme, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this programme to characterise the behaviour of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer's low temperature ratings are based on methods that simulate this use. The seal materials tested in this programme with a fixture similar to a RAM cask closure, with the exception of silicone S613–6O, are not leak tight (1.0 × 10–7 std cm³.s^?1) at manufacturer's low temperature ratings.     

The 'Anti-Evidence' Approach to Tamper Detection

Abstract

Tamper-indicating seals are widely used for the transport, packaging and storage of nuclear material. Most existing seals operate under the same basic principle: once the seal is opened, information that the seal has indeed been unsealed is stored in or on the seal until such time as the seal can be inspected. This stored evidence of tampering, however, is often easy to hide or erase. A better approach, discussed theoretically in this paper, is to store information when the seal is first installed that tampering has not been detected; this anti-evidence is then erased once the seal is opened. Such anti-evidence seals may provide better security. They also have a number of potentially useful attributes, including an intrinsic check on whether the seal was actually inspected for evidence of tampering ('anti-gundecking').     

Overall Reaction Rate Analysis of Ion-Exchange Resins Incineration by Fluidized Bed

A kinetic study on the incineration of ion-exchange resins was conducted using fluidized bed. In the experiment, cation or anion exchange resins with known quantities were fed into the fluidized bed maintained at a constant temperature from 550 to 750°C. The apparent reaction rate constants k _ap could be evaluated by the time for completion of combustion derived from the continuous measurement of CO₂ concentration in the off-gas.

It was confirmed that the reaction of the ion-exchange resins proceeded with the shrinking particle model forming no solid product layer and the rate of disappearance of the resins could be expressed by the surface chemical reaction.

Most preferable conditions for incinerating the ion-exchange resins were found to be about 650°C for temperature and more than 4.91 x 10^?2 m-s^?1 for the air velocities at fluidized bed u _B . Also, k _ap for cation and anion exchange resins were found to be 1.25x10^?2 and 1.51x10^?2 s^?1, respectively, at 650°C and u _B of 5.45x10^?2m.s^?1.     

Ageing of metallic gaskets for spent fuel casks: Century-long life forecast from 25,000-h-long experiments

An experimental programme is being carried out that aims at quantifying the relaxation of four types of metallic HELICOFLEX^® seals during their use in spent nuclear fuel storage casks. Two types of lining are taken into account: aluminium and silver. Tests longer than 10,000 h are implemented only for silver. For each type of lining, two different section diameters are investigated. The work aims at evaluating the minimum residual linear load that can be guaranteed for a seal after a particular time of relaxation. This relaxation depends on the evolution of the seal temperature with time. Therefore, holds of seals tightened between two flanges have been performed at several constant temperatures, including 100 and 200 °C. Residual load and ‘useful’ recovery have been measured after the holds. Results are interpreted according to two methods: a time extrapolation, and a time–temperature equivalence parameter. Both methods are based on linear relationships and are assessed through a statistical analysis (calculation of scatter) which is also used to determine a minimum guaranteed residual load. Finite element simulations of the relaxation of a seal have also been performed in order to justify qualitatively that the time extrapolation method is safe. For silver lining seals, the use of a time–temperature equivalence parameter equal to T (11 + log₁₀ (t)) appears justified and this enables us to assess the maximum temperature at which seals can be ‘safely’ used ‘up to a century’.Using the available ageing results (longest holds: 25,000 h), and the proposed prediction method, it can be proven that the two types of silver lining seals which are evaluated will retain a residual linear load of at least 100 N mm⁻¹ of seal perimeter after one century of use in a cask, if the initial temperature of the seal after closing the cask is less than or equal to 100 °C.     

Helium irradiation of Ni-(Zr or Nb) metallic glasses: Blistering,flaking and bubble formation

NiZr and NiNb are potential metallic glasses for nuclear applications. We have studied blistering, flaking and bubble formation in Ni₆₄Zr₃₆, Ni₃₃Zr₆₇ and Ni₆₀Nb₄₀ glasses under helium ion bombardment at room temperature. The effect of projectile energy (50–150 keV), total dose (0.01–10 × 10¹⁸ ions/cm²), dose rate (10–100 μA/cm²) and thermal crystallization on critical dose for blistering and/or flaking, average blister diameter and development of surface topography was examined. The surface damage effects in metallic glasses were in general similar to those for crystalline materials with the notable exception that the critical dose values are higher by about 50–100% for metallic glasses. TEM investigations revealed irradiation induced partial crystallization of Ni₃₃Zr₆₇ glass. Metallic glasses Ni₆₄Zr₃₆ and Ni₆₀Nb₄₀ were stable under helium irradiation and have also shown very high resistance against blistering and/or flaking. The possible mechanisms of helium trapping in metallic glasses are also discussed.     

Neutron Irradiation Effects on Thermally Activated Process of Tensile Properties of Titanium

This paper deals with the relationship between mechanical properties and irradiation, effects for titanium irradiated to fast neutron fluxes. The neutron fluences applied are 6.9×10¹⁸, 8.6 × 10¹⁸ and 3.0 × 10¹⁹ n/cm². Tensile deformation is carried out over the temperature range of 77–about 600°K retaining the strain rate constant on one hand and changing the strain rate by a factor of about 5 and 10 on the other.

The fluence (φ) dependence of the yield stress at room temperature for an athermal component of the stress, σμ is greater than that for a thermal component σ* which does not change remarkably after irradiation. Their increments Δσ, Δμ and Δ σ* are proportional toσ ^1/3, σ^1/2σ^1/4 and, respectively.

The relationship between activation volume V* and effective shear stress τ* is investigated for both the unirradiated and irradiated specimens. In terms of the τ*/τ*₀ analysis (τ*_o is the value of τ* at T = 0°K), V* shows a tendency to decrease with increase in neutron fluence.

Irradiation defects observable by electron microscopy seem to be related to the athermal activation stress (σ_u) and those too small to be observed by electron microscopy to the thermal activation stress. The yield stress in the thermal activation can be given by Conrad's formula. The activation energy H₀ shows a constant value of about 1.8 eV irrespective of the neutron fluence applied. This value is 0.3–0.4eV higher than that for unirradiated specimens.     

Modelling of release of particulate material from transport containers

Abstract

The Nuclear Decommissioning Authority (NDA) is developing a family of Standard Waste Transport Containers (SWTCs) for the transport of unshielded intermediate level radioactive waste packages. The SWTCs are shielded transport containers designed to carry different types of waste packages. The combination of the SWTC and the waste package is required to meet the regulatory requirements for Type B packages. One such requirement relates to the containment of the radioactive contents, with the IAEA Transport Regulations specifying release limits for normal and accident conditions of transport. In the impact tests representing accident conditions of transport, the waste package will experience significant damage and radioactive material will be released into the SWTC cavity. It is therefore necessary to determine how much of this material will be released from the cavity to the external environment past the SWTC seals. Typical assessments use the approach of assuming that the material will be evenly distributed within the cavity volume and then determining the rate at which gas will be released from the cavity, with the volume of radioactive material released with the gas based on the concentration of the material within the cavity gas. This is a pessimistic approach as various deposition processes would reduce the concentration of gas-borne particulate material and hence reduce their release rate from the SWTC. This paper assesses these physical processes that control the release rate and develops a conservative methodology for calculating the particulate releases from the SWTC lid and valve seals under normal and accident conditions of transport, in particular:

a) the flows within the SWTC cavity, especially those near the cavity walls;

b) the aerodynamic forces necessary to detach small particles from the cavity surface and suspend them into the cavity volume;

c) the adhesive forces holding contaminant particles on the surface of a waste package;

d) the breakup of waste material upon impact that will determines the volume fraction and size distribution of fine particulate released into the cavity.

Three mechanisms are specifically modelled, namely Brownian agglomeration, Brownian diffusion and gravitational settling, since they are the dominant processes that lead to deposition within the cavity and the easiest to calculate with much less uncertainty than the other deposition processes. Calculations of releases under normal conditions of transport concentrate on estimating the detachment of any waste package surface contamination by inertial and aerodynamic forces and show that very little of any contamination removed from the waste package surface would be released from the SWTC. Under accident conditions of transport, results are presented for the fraction released from the SWTC to the environment as a function of the volume fraction of the waste package contents released as fine particulate matter into the SWTC cavity. These show that for typical release fractions of 10^-6 to 10^-8 for the release of radioactive material from waste packages into the SWTC cavity, the release fraction of the waste package inventory from the SWTC of typically 10^-9 to 10^-10. Hence, the effective decontamination factor provided by the SWTC is 10² to 10³. Whilst this analysis has been carried out specifically for the SWTC carrying waste packages, it is applicable to other arrangements and its use would reduce the high degree of pessimism used in typical containment assessments, whilst still giving conservative results.     

Dependence of vacancy concentration on morphology of helium bubbles in oxide ceramics

Since the formation of helium bubbles degrades swelling property and thermal conductivity of minor actinide-containing mixed oxide (MA-MOX) fuel, it is essential to understand the conditions of the bubble formation. In order to examine the dependence of vacancy concentration on morphology of helium bubbles, helium was infused into (Zr,Fe)O_2?x. The oxygen vacancy concentration was controlled by addition of solute Fe³⁺ into ZrO₂. Helium was infused by hot isostatic pressing. The helium-infused specimens were observed by field emission scanning electron microscopy (FE-SEM) and field emission transmission electron microscopy (FE-TEM). In addition, X-ray fluorescence, X-ray diffraction analysis, conversion electron yield–X-ray absorption near-edge structure and FE-SEM/EDX (energy dispersive X-ray) analyses were also made to interpret the results of microstructure observations. As a result of the helium infusion treatment, numerous 0.5–10 nm bubbles were observed and its number density clearly depended on oxygen vacancy concentration. On the other hand, sizes of the helium nano-bubbles in all specimens were almost constant.     

Helium solubility and behaviour in uranium dioxide

A set of devices was developed in order to infuse UO₂ disks with helium, at high temperature and pressure, to measure the helium infused quantity and from these data to calculate the helium solubility in the UO₂ matrix. Samples of UO₂ single crystal and UO₂ polycrystal were infused at a temperature of 1473 and 1743 K in a helium atmosphere ranging between 50 and 100 MPa. These samples were then annealed and the helium released was measured with a mass spectrometer. From the obtained spectra it was possible to give an interpretation of the helium release mechanism and to calculate its solubility in the UO₂ lattice in these specific thermodynamic conditions. Additionally to the helium solubility measurement from infused samples, a 37 years old sample of ²³⁸PuO₂, retrieved from an old ²⁴²Cm radioisotope thermoelectric generator (RTG), containing radiogenic helium, was also measured to widen perspectives of this kind of measurements to damaged sample more representative of spent fuel.     

秦山核电厂安全壳系统B、C类密封性试验

叙述了秦山核电厂安全壳系统B、C类密封性能试验概况,主要包括试验范围、泄漏率分配、试验结果和总体评价等。     

Evaluation Method of Gas Leakage Rate from Transportation Casks of Radioactive Materials

Abstract

A sealing function is essential for transportation casks of radioactive materials in order to prevent radioactive materials from being released into the environment. In the safety analysis report, the release rate of radioactive materials into the environment is evaluated by using the method specified in ANSI N14.5 1977 and 1987. However, there are still problems of the evaluation method specified in ANSI N14.5 as follows:

(1) Gas leakage rate is discontinuous according to flow pattern.

(2) The gas compressibility is not considered for laminar flow.

(3) The units to indicate the gas leakage rate are not universal.

Therefore, in this paper, leakage rates from very narrow capillary tubes and orifices were investigated experimentally using He gas to obtain fundamental data for choked flow, non- choked free expansion flow and laminar flow with gas expansion. Next, an evaluation method of leakage rate of 10^?4 to 0.5 cm³/s was established. A simplified evaluation method was proposed for the leakage rates of 10^?4 to 10^?2cm³/s related closely to the sealing performance of the casks. Finally, the points to which attention should be paid for using the correlations specified in ANSI N14.5 were discussed.