Stress effect on the void swelling incubation period

The dose dependence of the swelling of solution-annealed Type 316 stainless steel irradiated as fuel pin cladding is compared with the behavior of irradiated stress-free samples. It is found that the onset of steady-state swelling occurs at a considerably lower fluence level for fuel pin cladding and a stress effect on the swelling incubation period is postulated to explain this result. A constituitive equation for this stress effect is formulated and a technique for using this constituitive equation with time-dependent stress histories is described. The stress effect on the incubation period is used in the calculation of the axial swelling profile of a fuel pin with 20% cold-worked Type 316 stainless steel cladding and good agreement with the measured swelling profile is obtained.     

Development of Long Life FBR Fuels with Particular Emphasis on Cladding Material Improvement and Fuel Fabrication

The development of FBR fuel systems with high reliability and long in-core residence capability is required to make the fast reactor economically competitive with other electrical energy sources. PNC program of fuels and materials development has been primarily focused on mixed uranium/plutonium oxide (MOX) fuel with cold-worked 316 stainless steel for the past 20 years. Modified 316 stainless steel with excellent swelling resistance and high creep rupture strength was obtained for cladding and duct of the fast prototype reactor MONJU. Advanced austenitic alloys and high strength ferritic alloys are also being investigated for high burnup fuel assemblies of a long life core in large scale FBRs.

In MOX fuel fabrication technology, extensive progress has been achieved during driver fuel fabrication for the experimental reactor JOYO. A new MOX production facility PFPF has been completed with fully automatic and remote handling systems. This facility serves for MONJU core fuel production. The improvement of fuel fabrication technologies promotes cost reduction, safety operation and security from a physical protection standpoint.     

Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 10²⁶ n/m² as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.     

Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.     

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.     

Swelling Behaviors in a Fuel Assembly for the Wrapping Wire and Duct Made of Modified 316 Austenitic Stainless Steel

Swelling behaviors in the wrapping wire and duct made of modified type 316 austenitic stainless steel were investigated in a fuel assembly irradiated in a fast breeder reactor. The temperature dependence of volumetric swelling was measured in the wrapping wire and the duct, and the peak temperatures of swelling were evaluated. The void distribution in the material was measured by microstructure observation with electron microscopy, and it was found that the voids prefentially grew near the surface. This phenomenon seemed to be caused by a surface effect on the neutron-irradiated materials.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Effects of Fast Reactor Irradiation Conditions on Tensile and Transient Burst Properties of Ferritic/Martensitic Steel Claddings

The effects of fast neutron irradiation conditions have been investigated by focusing on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated at temperatures from 693 to 833K to 42.5 dpa (displacement per atom) in the experimental fast reactor JOYO using the PFB090 fuel test subassembly. Post-irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for the PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. On the other hand, the strength of the HT9M cladding tended to shift to lower values than those of as-received specimens. The differences in tensile and transient burst strengths between the two claddings were attributed to martensite structural stability which was related to the stable solid solution elements.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

An approach for evaluating fuel system applied in innovative nuclear energy system

One of the important issues in the study of Innovative Nuclear Energy Systems (INES) is the integrity of the fuel system applied. An approach of evaluating fuel system integrity is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes controlling fuel life were reviewed and fuel integrity was analyzed and compared with the failure criteria.Metal and nitride fuels with austenitic and ferritic cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for INES, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic-martensitic steel (PNC-FMS) and oxide dispersion strengthen steel (PNC-ODS).The analytical result showed that fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In the case of ferritic steel, on the other hand, the fuel life is controlled by cladding creep rupture. The lifetime evaluated here is no more than 200 GWd/t, which is still lower than the target value 400 GWd/t burnup. Possible measures to extend metal fuel lifetime may be reducing fuel smear density and ventilating fission gas in the plenum.     

Voids resulting from fast neutron irradiation of a stabilized stainless steel

A 16 Cr-13 Ni-niobium stabilized stainless steel (Type 1.4988) was irradiated as fuel pin cladding in the DFR reactor. The irradiation temperatures ranged from 300 to 650 °C. Neutron fluences from 3.3 to 4.0 n/cm²$\text{(E > 0.1 MeV)}$ were achieved. Swelling behavior was studied in this material by transmission electron microscopy, immersion density determinations and diametral change measurements.Void formation was observed in the temperature range 360–610 °C. Void concentration values at lower temperatures (< 480 °C) are comparable to those cited for the unstabilized stainless steels AISI 304 and 316, however, they decrease more strongly with increasing irradiation temperatures. The mean and maximum diameters show a pronounced maximum at about 530 °C. The decrease of the void diameters at higher temperatures can be explained by the nature of the cavities observed. Annealing experiments with specimens irradiated at 610 °C have revealed a bubble or a mixed void-bubble character for these cavities.Swelling in type 1.4988 was found to be lower than that reported previously for types 304 and 316 at comparable irradiation and material conditions. This is especially pronounced at higher temperature. Diameter changes of the pin at $\text{T >- 610 °C}$ cannot be explained by void-formation.     

Impact of Material Configuration Behind First Wall on Helium Formation in In-Vessel Components of Fusion Reactor

The results of neutron transport calculations of the He formation based on the JENDL gas-production cross section file are discussed for some metals and alloys, namely ²⁷A1, Ti, ⁵¹V, Cr, ⁵⁵Mn, Fe, Ni, Zr, Mo, austenitic stainless steel (Ti modified 316 SS: PCA), Ni-base alloy (Inconel 625), ferritic steel (Fe-11Cr-1Mo: HT-9), Ti-base alloy (Ti-6A1-4V) and V-base alloy (V-5Cr-5Ti). Impacts of the two shields having the steel-rich and the H₂O-rich compositions and the two blankets having the Li₂O/Be-base and the liquid Li/Be-base compositions on the He formation rate in the above-mentioned metals and alloys are discussed. The relation between the He formation rate and the fast neutron flux (14.1 MeV>E>0.1 MeV) is investigated. The decrease of He formation at any distance Δ from the first wall more than Δ_as, the distance where the shape of neutron spectrum reaches its asymptotic form, is modelled by the simple formula based on the exponential dependence, as those reported so far for the fast neutron flux and the displacement damage rate.     

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail.     

碳钢对核主泵用奥氏体不锈钢的污染研究

奥氏体不锈钢在加工、运输和装配过程中如果与碳钢直接接触,就会被碳钢污染,而导致奥氏体不锈钢耐蚀性能的改变。众所周知,核主泵用奥氏体不锈钢对耐蚀性有着非常严格的要求,本文以Z2CN18-10核主泵用奥氏体不锈钢为例,通过FeCl3腐蚀试验和电化学方法测试了被碳钢污染后其耐腐蚀性能的变化。试验结果表明:附着在不锈钢表面的碳钢对其长期总体腐蚀速率影响不大;嵌入式的碳钢颗粒会显著降低奥氏体不锈钢的点蚀电位,增大发生点蚀的倾向;硝酸钝化可部分抵消被污染不锈钢点蚀电位的降低,但该值仍远低于同样经过硝酸钝化,而未被污染的不锈钢的点蚀电位。此外,还针对碳钢污染对核电站辐射场的影响和对燃料包壳热传导效率的影响进行了讨论。     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

Evaluation of neutron economical effect of new cladding materials in light water reactors

The present paper aims to contribute from a neutronic aspect to activities for new cladding material development for light water reactors (LWRs) that can reduce the risk of hydrogen gas explosion. Iron (Fe), nickel (Ni), titanium (Ti), niobium (Nb) and vanadium (V) are selected as possible component elements to cover a variety of new cladding materials for LWRs. The effect of larger thermal absorption cross sections of these elements than those of zirconium (Zr), together with those of silicon carbide (SiC), on the neutron economy in LWRs is evaluated by performing pin cell burnup calculations for a conventional pressurized water reactor (PWR), a low-moderation high-burnup LWR (LM-LWR) and a high-moderation high-burnup LWR (HM-LWR). As can be anticipated from the thermal cross sections, SiC has excellent neutron economy. The materials other than SiC largely decrease discharge burnup for all three types of LWRs in comparison with Zircaloy-4. Among such elements of larger thermal absorption cross section, Nb has neutron economical advantage over the other materials except SiC in softer neutron spectrum reactors such as HM-LWR in which the atomic number ratio of hydrogen to heavy metal is 6. In conventional LWRs, stainless steel of low Ni contents has the neutron economic advantage as well as Nb for cladding material. The results of the calculations are summarized for the purpose to provide reference data for new cladding material development studies, in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding.     

Ti-5%Ta钛合金在乏燃料模拟溶解液中的腐蚀行为

000Cr25Ni20超低碳奥氏体不锈钢目前作为乏燃料后处理中溶解器设备的材料，在后处理的溶解工况下腐蚀严重．本文通过均匀腐蚀模拟试验对Ti-5％Ta钛合金和000Cr25Ni20奥氏体不锈钢在动力堆乏燃料模拟溶解液中的均匀腐蚀行为进行了研究：研究发现Ti-5％Ta钛合金的抗腐蚀性能远优于000Cr25Ni20奥氏体不锈钢。原因是Ti-5％Ta钛合金试样的表面形成了致密的氧化膜，阻止了腐蚀的进一步发展，而在000Cr25Ni20奥氏体不锈钢试样的表面未发现氧化膜的存在。     

Properties of Plutonium Dioxide as Nuclear Fuel

Generalized structural, thermodynamic, thermophysical, and strength characteristics as well as data on the dimensional stability, compatibility, and fragment gas release under irradiation of plutonium dioxide, used as a compact nuclear fuel for fuel elements in fast research reactors, are presented. Reliable operation of reactors with fuel elements based on plutonium dioxide confirmed that the properties of the fuel ceramic of fuel elements with austenitic stainless steel cladding have been determined with a high degree of reliability. 7 figures, 6 tables, 23 references.     

New approach to evaluate lattice expansion of light water reactor fuel elements on criticality safety of transport packages under impact accidents

Abstract

In order to safely transport packages containing light water reactor fuel assemblies, it is essential to maintain the fuel assemblies in a subcritical state in accidents during transport. To evaluate nuclear criticality safety, an estimator is required to determine an absolutely safe level based not only on hypothetical accidents but also on practical accident levels which, to some extent, are based on actual accidents. The purpose of the present study is to suggest the arrangement of the deformation range of the fuel assembly after an actual accident, and to obtain the maximum value of the neutron effective multiplication factor based on the criticality safety assessment for the transport cask. In the present study, two kinds of criticality calculations for the package were considered: large scale pin pitch shift and small scale pin pitch shift. For the large scale pin pitch shift, a parameter which determines the location of each fuel pin which constitutes the fuel assembly was introduced so that the criticality calculation for the fuel assembly with non-uniform lattice pitch can be performed parametrically. The result of the criticality calculation using the parameter made it clear that the fuel pin pitch is sensitive to the neutron reactivity because each of the fuel pin pitches is related to a ratio of the fissile to the moderator, and that the relationship of the ratio to the neutron reactivity depends on the type of the fuel assembly involved, i.e. the type of a nuclear reactor in which a fuel assembly is used. For the small scale pin pitch shift, the study focused on the small displacement of each fuel pin. The small displacement of each fuel pin pitch can be described probabilistically using the stochastic geometry routine in MCNP code. Using the scheme in combination with the scheme for the large scale pin pitch shift, the maximum value of the neutron effective multiplication factor of the package after an accident can be obtained. This scheme is useful to determine the maximum neutron effective multiplication factor for the criticality safety evaluation.     

Physics and safety studies of a compact ADS design loaded with nitride fuel

We have performed transient analysis of a nitride fueled and heavy liquid metal cooled accelerator driven system, in which the pitch-to-diameter ratio of pin lattice was set at 1.26. Unprotected loss-of-flow, unprotected transient-over-power and protected loss-of-heat-sink transients were simulated in a geometrical model of the suggested ADS design, using safety parameters obtained with the MCNP/MCNPX code.The simulations indicated that the suggested ADS design could survive the full set of transients, thanks to the introduction of the austenitic 15/15Ti stainless steel as fuel cladding material.Thus, the viability of an ADS design with small pin pitch and concomitant high proton source efficiency could be confirmed.