Study on optimum zirconium liner thickness for barrier fuel

The superiority of barrier fuel over the non-barrier fuel is verified in this paper. Based on the strain energy density theory, a thorough study on the zirconium liner thickness of barrier fuel cladding for various burnup conditions is established. It is found that the presence of zirconium liner does substantially reduce the local strain energy density available for failure initiation and also enhance the system stability. Since an extensive increase in the zirconium liner thickness does not further improve the fracture strength as well as failure stability, an optimal thickness of the zirconium liner is then determined in the present study.     

Behavior of metallic precipitate in an irradiated simulated fuel

A simulated fuel specimen which was irradiated at the HANARO research reactor up to 3300 MWd/tU of a burn-up at the condition of 36 kW/m of a maximum linear power was studied by a shielded EPMA (Electron Probe Micro-Analyzer). In order to obtain an accurate analysis results, chemical and EPMA analyses were also performed on un-irradiated fresh simulated fuel, the results of which were compared with those of the irradiated simulated fuel. This study concentrated on the metallic precipitates of the irradiated simulated fuel specimen which contained lots of fission products. Among the several properties of the metallic precipitate, its size and composition were investigated. A large metallic inclusion was also observed in the irradiated simulated fuel, from which X-ray photographs were taken to analyze its properties.     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U³⁺, Pu³⁺, Np³⁺, Zr²⁺ or Zr⁴⁺, Am³⁺, Nd³⁺, Y³⁺, Ce³⁺ and Gd³⁺ chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (−1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium.     

Migration of minor actinides and lanthanides in fast reactor metallic fuel

Minor actinides (MA) and lanthanides (LA) migration to the cladding in fast reactor metallic fuel is a concern because of their reaction with the cladding material. MA and LA migration test results are analyzed and reactions with the cladding are characterized. Remedies for reducing MA and LA migration and reaction with cladding are reviewed. A possible method is proposed that includes the addition of a compound forming element such as indium or thallium in fuel. These elements preferentially form stable compounds with MA and LA and should therefore reduce migration of MA and LA. As an intrinsic solution to the issue, the feasibility of the use of U-Pu-Mo alloy as fuel is studied. Unlike U-Pu-Zr, U-Pu-Mo consists of a single phase at typical fuel operation temperatures, and should have negligible fuel constituent redistribution and reduce MA and LA migration.     

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

A study on reactivity insertion controlled LMR cores with metallic fuel

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity.

A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core.

The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.     

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.     

Performance analysis of coated plutonia particle fuel compact for radioisotope heater units

Coated plutonia particle fuel has been proposed recently for use in radioisotope power systems and radioisotope heater units for a variety of space missions requiring power levels from milliwatts to tens or even hundreds of watts. The ²³⁸PuO₂ fuel kernels are coated with a strong layer of ZrC designed to fully retain the helium gas generated by the radioactive decay of ²³⁸Pu. A recent investigation has concluded that helium retention in large-grain (200 μm) granular and polycrystalline fuel kernels is possible even at high-temperatures (>1700 K). Results of performance analysis showed that this fuel form could increase by 2.3–2.4 times the thermal power output of a light weight radioisotope heater unit. These figures are for a single-size (500 μm) particles compact, assuming 10% and 5% helium gas release respectively, and a fuel temperature of 1723 K, following 10 years of storage. A binary-size (300 and 1200 μm) particles compact increases the thermal power output of the RHU by an additional 15%.     

乏燃料储存与运输两用容器安全性能研究进展

乏燃料运输和储存两用容器具备乏燃料运输和储存两种功能,是乏燃料实现最终贮存和处置前的一种储运方式。本文介绍国际乏燃料储存与运输两用容器安全设计要求和安全验证实践经验,研究适合我国乏燃料储存与运输两用容器安全设计要求和安全验证要求,为我国乏燃料储存与运输安全提供参考。     

Electrorefining of metallic fuel with burn-up of ~7 at% in a LiCl-KCl melt

Electrorefining of irradiated metallic fuels (burn-up ~ 7 at%) in a LiCl-KCl melt at 773 K was successfully demonstrated: Actinides in the fuels were anodically dissolved in the melt. Both a selective U metal deposition on a solid cathode and a grouped recovery of actinides, U, Pu, Np, Am, and Cm, in a liquid Cd cathode were confirmed. The behavior of fission products, such as lanthanides, alkali metals, alkaline earth metals, and noble metals, were also investigated. It was found that the behaviors of actinides and fission products in the electrorefining of the fuels with ~ 7 at% burn-up were similar to those in electrorefining of fuels with ~ 2.5 at% burn-up.     

Stopping power of polymeric foils for swift heavy ions

The stopping power of polypropylene PP(C₃H₆) and Polyethylene naphthalate PEN (C₇H₅O₂) polymeric foils has been measured, using transmission technique, for Si, Cl and Ti ions covering the energy range 1.0–4.5 MeV/u. These measured stopping power values have been compared with the corresponding values generated from the widely used semi-empirical formulations and standard data tables. The applicability of these formulations and data tables, in the light of the experimental values, is highlighted.     

Measurement of the specific heat of Zr–40 wt%U metallic fuel

The specific heat capacities of un-irradiated and irradiated metallic Zr–40 wt%U fuel have been measured between 50 °C and 1000 °C with a differential scanning calorimetry. The irradiated fuels have three different burnup levels of 0.38, 0.70 and 0.92 g-fission product (FP)/cm³. The measured specific heat for the un-irradiated fuel is representative and consistent with the values estimated from the Neumann–Kopp rule. The irradiated fuels exhibited a complicated behavior of the heat capacities. The unique characteristics of the specific heat capacities can be explained by the recovery of radiation damage, the formation of fission gas bubbles and fission gas release, and a phase transition in the irradiated fuels. An examination of the microstructure revealed that multiple large bubbles were formed in the irradiated fuel during specific heat measurement. The measured specific heat is expected to enable us to estimate the stored energy in the metallic fuel during certain accident scenarios and to determine the thermal conductivity of zirconium–uranium metallic fuel.     

In-pile release behavior of metallic fission products in graphite materials of an htgr fuel assembly

Distribution of metallic fission products in the graphite sleeve and block of the fifth OGL-1 fuel assembly was measured by gamma spectrometry with lathe sectioning. Considerably large release fractions of long-lived fission products with smooth axial profiles were observed in the sleeve due to a large failure fraction of coated fuel particles accompanied with failed silicon carbide layers. Nevertheless, a key nuclide^110mAg, whose large release is suspected at increased burnups for low-enriched uranium fuels, was effectively retained within the graphite sleeve. The retention was also observed for¹²⁵Sb, ¹⁵⁴Eu and¹⁵⁵Eu up to a burnup of 3.2% fission per initial metal atom, but was limited for¹³⁴Cs and¹³⁷Cs at high sleeve-temperatures above 900°C. In-pile diffusion coefficients in IG-110 graphite have been evaluated for Cs, Ag and Sb; those for Cs are in reasonable agreement with available in-pile data.     

Modeling of fuel performance and metallic fission product release behavior during HTTR normal operating conditions

The computer codes PANAMA and FRESCO developed at the Research Center Jülich have been used for the prediction of fuel performance and fission product release behavior during the normal operation of the Japanese High-Temperature Engineering Test Reactor, HTTR. Basis for the calculations was the so-called ‘Standard HTTR Operation Plan’ with a nominal operation time of 660 efpd including a 110 efpd period with enhanced fuel temperatures. Fuel performance model calculations with the PANAMA code have shown that for the temperature distribution given, only a small additional failure fraction is expected. The diffusive release of metallic fission products from the fuel occurs mainly from the central core layers with the maximum temperatures whereas there is little contribution from the upper layer. Silver most easily escapes the fuel. The release data for strontium and cesium also reveal a significant fraction to originate from still intact particles. The comparison with the calculations obtained with the JAERI models has shown a good agreement for the release from the coated particles.     

Atomistic simulations on the interfacial interaction of metallic fuel and structural materials in SFRs - molecular dynamics model for Pu-Fe system

Interaction between metallic fuel and steel structures is one of the predominant phenomena in the progress of core disruptive accidents of Sodium-cooled fast reactor. In this study, the atomic diffusion across the interface between Pu and Fe was investigated by using molecular dynamics. The simulation was performed by using Modified Embedded Atom Method (MEAM). The interactions between plutonium and iron atoms were calculated by using the newly developed potential model determined so as to reproduce the material properties of PuFe₂ and Pu₆Fe. The material properties of the compounds predicted with the developed potential were in good agreement with the referenced data. The dissolution or melting at the interface between solid Fe and solid or liquid Pu were simulated by contacting semi-infinite slabs (or liquid layer) of them. Dissolution was observed for all the tested temperature conditions from 800 K up to 1700 K. The melting at the interface was also observed on the interface between solid Fe and PuFe₂ slabs at the temperature approximately 100 K below the melting temperature of PuFe₂ obtained based on the present model.     

Energy loss of Xe-ions in thin foils

The energy loss of ¹³²Xe-ions in C, Al, Ni, Ag, Lu, Au, Pb and Th foils was measured in the energy range from 0.1 to 5 MeV/u using the TOF-E method. The results are compared with previously published data and with the predictions of several computer codes. They include theoretical codes: PASS, CASP, semi-empirical programs: SRIM, LET and the Hubert table predictions.