Non destructive determination of PuO2 content in MOX fuel pins for fast reactors using Passive Gamma Scanning

Use of Passive Gamma Scanning for non destructive evaluation of PuO₂ content in mixed oxide (MOX) fuels for fast reactors is demonstrated. Experiments have been carried out on MOX fuel pins for the hybrid core of Fast Breeder Test Reactor having nominal PuO₂ content of 44% and MOX pins having nominal PuO₂ content of 21% for the Prototype Fast Breeder Reactor. A comparison of results obtained using a conventional NaI(Tl) detector and that using a through well shaped detector is also presented.     

New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

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.     

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.     

Use of Thorium-Plutonium MOX in the inner pins of CANDU fuel bundles

This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO₂ fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO₂. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

Rupture and fragmentation of pressurized pipes and fast reactor fuel pins

A model of axial crack propagation in a pressurized tube is developed which predicts the crack velocity and deformation geometry and the minimum driving pressure. Emphasis is placed upon the stability of propagation. The model also offers a criterion for the appearance of multiple cracks and subsequent fragmentation of the tube wall due to excessive axial bending strains. The model is applied to the rupture of gas pipelines, PWR coolant pipes and fast reactor fuel pins.     

Fuel-cladding mechanical interaction in fast breeder fuel pins; observations and analysis

The problems of fuel-cladding mechanical interaction are considered, and a survey is given of the causal processes in oxide fuel pins. Their importance is judged by relevant results from irradiation experiments in thermal and fast test reactors, and by corresponding modelling computations. It is demonstrated that critical cladding strain has to be expected only in case of large, fast rod power increase, primarily from reduced power to full power, and maybe also in case of serious cesium accumulation at the fuel/blanket interface. Steady-state fuel swelling does not make any considerable contribution to cladding strain by thermal creep or plastic flow.     

Proliferation-resistant fuel options for thermal and fast reactors avoiding neptunium production

Sustainable nuclear energy production requires reuse of spent nuclear fuel while avoiding its misuse. In the paper we assume that plutonium with sufficiently high content of the Pu-238 isotope (about 6% or more) and americium from spent nuclear fuel are proliferation-resistant. On the other hand, neptunium should be considered as material that is fissionable in a fast neutron spectrum and could be misused.We also assume that plutonium denatured by Pu-238 can be produced in nuclear reactors of, e.g. nuclear weapon states and used for fuel fabrication there or in multilateral reprocessing and re-fabrication centers as suggested by IAEA. Then the fabricated fuel can be utilized in nuclear reactors everywhere provided that the reactors may operate safely and the fuel remains proliferation-resistant after utilization. Options to meet these criteria are investigated in the paper for two reactor types: pressurized water reactors (PWRs) and fast reactors (FRs).In PWRs, the investigated fresh fuel compositions include denatured plutonium and depleted uranium mixed with a small amount of U-233, thorium and, optionally, with americium, presence of U-233 making the coolant void effect negative. In FRs, use of americium makes plutonium denatured, both for the burner (without fertile blanket) and breeder options. It is shown that the proposed design and fuel options are proliferation-resistant, the generation of neptunium being very low. Safety parameters are acceptable. Advanced aqueous or pyrochemical reprocessing for plutonium/thorium/uranium fuel and related fuel re-fabrication technology applying remote handling may become necessary to realize the considered fuel cycles.