Power-to-melts of uranium-plutonium oxide fuel pins at a beginning-of-life condition in the experimental fast reactor JOYO

Power-to-melts of uranium-plutonium oxide fuel pins at an initial startup condition were experimentally obtained from the B5D-2 test in the experimental fast reactor JOYO in Oarai Engineering Center. MCNP code calculations were combined with burnup measurements to determine linear heat rating of the test fuel pins. To identify the axial incipient melting positions corresponding to the power-to-melts, solidified grain morphology and molten fuel axial movements were characterized. Extensive observations on longitudinal ceramographs allowed classifying molten fuel settlements near bottom and top extents of axial fuel melting into three types. The power-to-melts depended slightly on fuel-to-cladding gap sizes and clearly on both oxygen-to-metal ratios and densities of fuel pellets. These dependencies resulted from the fuel pellet cracking and relocation behavior, which fairly improves heat transfers across the gaps. Also, the power-to-melt at the bottom position was higher than that at the top position due to an axial gradient of cladding temperatures in each fuel pin.     

Measurement of subassembly outlet coolant temperature in the JOYO experimental fast reactor

Thermocouple temperature sensors are installed above the central region of the core in the JOYO experimental fast reactor to monitor the outlet coolant temperature of 115 subassemblies. This paper summarizes the experimental temperature data obtained during initial 50 MWt operation of the reactor. Subassembly outlet coolant temperature distributions that were obtained under various power levels, different main cooling system flowrates, and unequal reactor inlet temperatures from the two cooling loops are described. In addition, coolant temperature and flowrate distributions at the subassembly outlet measured in a zero power experiment are presented.     

Axial strains in fuel cladding associated with creep and fast neutron irradiation

Commercially produced CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) Zircaloy-4 and Zr—2.5% Nb fuel cladding was biaxially creep-tested in the laboratory and in the WR-1 reactor. The axial strains measured have been interpreted, through a knowledge of the texture, to provide evidence supporting prismatic slip as the major process for axial contraction at high stresses and temperatures and that this is the same process which gives rise to axial lengthening of pressure tubes. At low stresses and temperatures, axial lengthening of fuel cladding appears to be associated with Coble creep in elongated and flattened grains.     

The fast neutron energy spectrum and the thermal neutron flux distribution in the experimental channel of the VVR reactor

Using threshold detectors of aluminum, indium, iron, titanium, zinc, silicon, magnesium, mercury, and sulfur, we have measured the fast-neutron energy distribution in the experimental vertical channel of the VVR-2 reactor, traveling in the immediate vicinity of the active zone. We have obtained the spectral composition of the fast neutrons in the center of the channel at the level of the middle of the active zone, the thermal neutron flux and its distribution in the channel. The activity of all detectors except the sulfur and copper detectors was determined from the -radiation. The obtained experimental data are compared with the results of the theoretical calculation.     

Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor

In order to investigate the effect of americium addition in MOX fuel on the irradiation behavior, the ‘Am-1’ program is being conducted in the experimental fast reactor Joyo. The Am-1 program consists of two short-term irradiation tests of 10 min and 24 h irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. This paper reports on the results of PIEs for Am-containing MOX fuel irradiated for 10 min. MOX fuel pellets containing 3% or 5% Am were fabricated in a shielded air-tight hot cell using a remote handling technique. The oxygen to metal ratio (O/M) of these fuel pellets was 1.98. They were irradiated at peak linear heating rate of about 43 kW m⁻¹. Focus was being placed on migration behavior of Am during the irradiation. The ceramography results showed that structural changes such as lenticular pores and a central void occurred early, within the brief 10 min of irradiation. The results of electron probe microanalysis revealed that the concentration of Am increased in the vicinity of the central void.     

Behavior of Si impurity in Np-Am-MOX fuel irradiated in the experimental fast reactor Joyo

The irradiation behavior of uranium-plutonium mixed oxide fuels containing a large amount of silicon impurity was examined by post-irradiation examination. Influences of Si impurity on fuel restructuring and cladding attack were investigated in detail. Si impurity, along with Am, Pu and O were transported by spherical pores and cylindrical tubular pores to the fuel center during fuel restructuring of the Np-Am-MOX fuel, where a eutectic reaction of fuel and Si-rich inclusions occurred. After fuel restructuring of the Np-Am-MOX fuel, Si-rich inclusions without fuel constituents were agglomerated at fuel crack openings where shallow attacks on the inner wall of the cladding were seen. Such shallow attacks on the inner wall of the cladding were likewise observed near the location of fuel cracks in long-term steady-state irradiated MOX fuels. Evidence of these shallow attacks on the inner wall of the cladding remained after fuel restructuring in normal MOX fuel. However, grain boundary corrosion of the cladding inner wall at the opening of the fuel cracks was selective and was marked in MOX fuel at higher oxygen potential by the release of reactive fission products such as Cs and Te in comparison with other regions of cladding wall.     

Power effect of reactivity in fast power reactor with allowance for behavior of fuel under irradiation

Conclusions The proposed computational approaches for determining the thermal conditions of the fuel elements with allowance for their operating history have made it possible to bring closer together the calculated and experimental values of the power effect of reactivity and its dependence on the power level in the BN-350.Analysis of the causes of the divergence of the experimental and calculated values of the power effect of reactivity showed that these divergences, as well as changes in the effect during operation, are due mainly to the thermal conditions of the fuel elements. The principal indeterminacy in the calculation of the thermal conditions of the fuel elements is due to the contact thermal conductivity and its variation with the operating conditions of the fuel elements in the reactor. In order to eliminate this indeterminacy it is necessary to carry out experimental investigations to ascertain the contact thermal conductivity as well as to refine the physicomechanical properties of the fuel (thermal conductivity, ultimate strength etc.) as a function of the burn-up and temperature.Translated from Atomnaya Énergiya, Vol. 47, No. 3, pp. 157–161, September, 1979.     

The irradiation test of inert-matrix fuel in comparison to uranium plutonium mixed oxide fuel at the halden reactor

Conservative modelling for pin layout shows that the relatively low thermal conductivity of Inert-Matrix Fuel (IMF) causes higher temperatures and therefore higher fission gas release than in uranium plutonium mixed oxide (MOX). According to neutronic calculations, performance differences will also arise from different evolutions of the respective radial power and burnup distributions. Modelling of these effects as well as a 10% greater production of Xe in the thermal spectrum of the Halden reactor is well within the capabilities of appropriate codes. Some of the data and models used for the pre-calculations are preliminary and will be revised after the first experimental data have become available.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test

Studies in support of the assessment of aging structural materials in pressurized water reactors are being performed at the Paul Scherrer Institut. To that aim, a state-of-the-art methodology based on applying a CASMO-4/SIMULATE-3/MCNPX calculation scheme has been developed. In the frame of the methodology validation, an investigation is currently reported pertaining to the sensitivity of the calculated results, for a specific reactor pressure vessel scraping test, to the nuclear data used with the Monte Carlo code. Thus, the MCNPX-2.4.0 calculations have been carried out using three different data libraries, based on JEF-2.2, ENDF/B-VI.8 and JENDL-3.3 evaluations, respectively.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

中国实验快堆泵支承冷却系统温度场分析

中国实验快堆一回路泵支承套筒是承重设备,位于高温的热钠池中.为了限制套筒和套简内部冷钠腔室的钠温度,维持主泵正常工作温度,设置钠泵支承冷却通道.利用计算流体动力学技术(CFD),对泵支承冷却系统进行三维模拟,通过对泵支承冷却系统冷却流道和支承结构的数值传热分析,得到了该系统的温度场分布情况,验证了泵支承冷却系统的冷却能力.     

Metallic fast reactor fuel fabrication for the global nuclear energy partnership

Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.