Thinking outside the box: options for transport of high burnup spent fuel

Abstract

The current uncertainty surrounding the licensing and eventual opening of a long term geologic repository for the nation’s civilian and defense spent nuclear fuel and high level radioactive waste has shifted the window for the length of time spent fuel could be stored to periods of time significantly longer than the current licensing period of 40 years for dry storage. An alternative approach may be needed to the licensing of high burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.     

压水堆核电厂乏燃料组件源项计算分析

核燃料贮存、运输以及后处理过程中的安全是构成核与辐射安全的重要内容,为保证安全性,提高运输经济性,减小后处理厂对环境的排放,须获得乏燃料组件的包络源项,因此,采用ORIGEN-ARP程序分析组件运行历史、初始富集度、燃耗深度等参数对源项的影响。运行历史在卸料初期对源项略有影响,可采用合适的保守因子予以包络,在冷却一定时间后,其影响可忽略不计;初始富集度、燃耗深度均不同的组件须经对比计算以获得包络源项。计算表明：在目前核电厂乏燃料组件中,²³⁵U初始富集度为4.45%、燃耗深度为55 GW•d/tU的AFA-3G型组件源项是包络的,可作为乏燃料水池、运输容器设计,以及后处理厂排放源项分析的初始源项。     

The performance of defected spent LWR fuel rods in inert and dry air storage atmospheres

A testing program using eight commercial PWR and BWR spent fuel rods was conducted to investigate their long-term stability under a variety of possible dry storage conditions. The objective of this project is to provide the Nuclear Regulatory Commission (NRC) with the information to confirm or establish spent-fuel, dry storage licensing positions regarding long-term, low-temperature ( <523 K) spent fuel rod behavior during dry storage, and for radioactive contamination arising from spallation of cladding crud. Until now, the testing program has included three interim nondestructive examinations and one destructive examination. This paper presents the results of the third examination conducted to determine any degradation in eight fuel rods after being subjected to 13168 h at temperature. During this examination, visual observations, diametrical measurements, and isotopic analysis of smears were used to assess the fuel rod behavior and particulate release.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

The OECD Halden reactor project fuels testing programme: methods, selected results and plans

The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet–clad interaction. Relevant burnup levels (>50 MWd kg⁻¹ U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000–2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations.     

Effect of Pu-Rich Agglomerate in MOX Fuel on a Lattice Calculation

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17 × 17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

Finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel storage cask

A spent fuel storage cask is required to prove the safety of its canister under a hypothetical accidental drop condition which means that the canister is assumed to be free dropped on to a pad of the storage cask during the loading of the canister into a storage cask. In this paper, finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel dry storage cask were carried out to improve the structural integrity of the canister under a hypothetical accidental drop condition. The pad of the storage cask was originally designed as cylindrical steel structure filled with concrete. The pad was modified by using the structure composed of steel and polyurethane-foam instead of the quarter of the upper concrete as an impact limiter. The effects of the shape and the thickness of the steel structure and the density of the polyurethane-foam which was used in between steel structures were studied. As the optimized pad of a spent fuel dry storage cask, the quarter of the upper concrete was replaced with 12 mm thick circular steel structure and polyurethane-foam whose density was 85 kg/m³. The drop tests of a 1/3 scale model for the canister on to the original pad and the optimized pad were conducted. The effect of the pad structure was evaluated from the drop tests. The optimized pad has a greater shock-absorbing effect than the original pad. In order to verify the analysis results, strains and accelerations in the time domain by the analytical methods were compared with those by a test. The numerical method of simulating the free drop test for a dry storage cask was verified and the numerical results were found to be reliable.     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

On the influence of spatial discretization in LWR-burnup calculations with HELIOS 1.9 – Part II: Mixed oxide (MOX) fuel

Cell and burnup calculations are the basis for all deterministic static and transient 3D full core calculations for different operational states of the reactor. The arising differences in the integral transport solution (neutron flux and k_inf) for different discretization strategies during the burnup of mixed oxide (MOX) fuel due to different spatial discretization are demonstrated. The influence of different discretization strategies on the calculation of homogenized few group cross-sections is investigated. The influence of the discretization strategies on the calculation time is evaluated.