A non destructive technique for determination of U and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U²³³O₂ and PuO₂ have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO_2, it was difficult to monitor the percentage of U²³³O₂ through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U²³³O₂ and PuO₂ and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

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.     

Criticality of the thorium burnup in equilibrium state

The effective neutron multiplication factor (k_eff) as a function of burnup for different volume coolant (CoR) and fuel (FR) to cell ratio is presented. Additionally the Conversion Ratio (CR) of Th-232 to U-233, concentration of U-233, fissile and fission products calculation as a function of burnup are presented. The assembly is a critical reactor which makes volumes of coolant and fuel changes possible. In addition, an analytical model of calculation of k_eff as a function of U-233 and a poison concentration in equilibrium state are presented. One can achieve the criticality of Thorium Breeder Reactor (TBR) for enough high average neutron energy which one can obtain in Fast Breeder Reactor (FBR) only. The maximal value of CR and burnup for case of k_eff ≥ 1 achieves 1.4 and 360 GWd/MTU, correspondently. The calculations were done with a MCNPX 2.7 code using F₂Be, Na and Pb coolants.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

Plutonium dioxide (PuO₂) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO₂ at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO₂ were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO₂ were investigated up to 1500 K.     

A feasibility study of ferro-boron as in-core shield material in fast breeder reactors

Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder Reactor (PFBR), a 500 MWe which is sodium cooled, pool type, mixed oxide (MOX) fuelled reactor, which is under construction at Kalpakkam, India. It is shown through 2D transport calculations, carried out using 175 neutron multigroup cross-sections, that this low cost material as shield is capable of satisfying the radiological safety criteria as well as the shields in the reference case. The secondary sodium activity and dose in steam generator building are marginally lower than the reference case. The total shield material weight will be lower by about 50 tonnes and the material cost lower by a factor 5 as compared to PFBR shields comprising of stainless steel and B₄C.     

Comparison of FBTR hybrid core and Mark-II (carbide) core design and neutron irradiation in invessel shields

Fast Breeder Test Reactor (FBTR) in Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is presently operating with a combination of Mark-I (PuC 70% and UC 30%) and Mark-II (PuC 55% and UC 45%) fuels. Recently it became necessary to review the type of fuel for the FBTR core due to the use of oxide fuel for the PFBR under construction at Kalpakkam. A full Mark-II core was earlier designed to take the reactor to higher power. An alternate strategy of having hybrid core with Mark-I fuel at the centre and MOX fuel at the periphery was also studied. MOX subassemblies have a fissile column length of 43 cm which is higher than Mark-I subassemblies (32 cm). The lower position of these subassemblies is closer to the grid plate and therefore there was a concern on the grid plate fluence below the MOX fuel. It is interesting to compare the dose rates to the grid plate in the two above design as the doses determine reactor life. Many 2-D R-Z transport calculations were carried out for FBTR including all regions of interest like borated graphite and carbon steel in the top shield and vault concrete to obtain 175 group neutron flux distribution in Vitamin J structure with IGC-S3 self shielded cross-section set. The radiation damage to grid plate below MOX part is found to be less than that below Mark-I part of hybrid core in spite of longer fissile column length. Similar behaviour is seen for helium production as well. Comparison showed that the difference in fluence in the grid plate region is less than 3% between the two core designs.     

铅冷快增殖堆物理和热工水力研究

对25MW电功率铅冷快增殖堆堆芯进行了物理和热工水力概算，并将计算结果与相同功率的钠冷快增殖堆的结果进行了分析比较。从初步概算的结果来看，铅冷快增殖堆是一种安全可行的快增殖堆堆型。     

Oxygen potential and thermal conductivity of (U, Pu) mixed oxides

(U, Pu) mixed oxides, (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 are being considered as fuels for the Prototype Fast Breeder Reactor (PFBR) in India. The use of urania-plutonia solid solutions in PFBR calls for accurate measurement of physicochemical properties of these materials. Hence, in the present study, oxygen potentials of (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 were measured over the temperature range 1073-1473 K covering an oxygen potential range of −550 to −300 kJ mol⁻¹ (O/M ratio from 1.96 to 2.000) by employing a H₂/H₂O gas equilibration technique followed by solid electrolyte EMFmeasurement. (U_1−yPu_y)O_2−x, with y = 0.40 is being used in the Fast Breeder Test Reactor (FBTR) in India to test the behaviour of fuels with high plutonium content. However, data on the oxygen potential as well as thermal conductivity of the mixed oxides with high plutonium content are scanty. Hence, the thermal diffusivity of (U_1−yPu_y)O₂, with y = 0.21, 0.28 and 0.40 was measured and the results of the measurements are reported.     

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.     

OASIS程序的开发与应用

全面描述了对来自法国原子能委员会 (简称CEA)的快堆系统安全分析程序OASIS的引进和开发工作 ,并在此基础上介绍了该程序在中国实验快堆 (ChinaExperimentalFastReactor,简称CEFR)初步安全分析报告中对主给水管道断裂事故的分析计算。     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

PFR fuel cladding transient test results and analysis

Fuel Cladding Transient Tests (FCTT) were performed on M316 cladding specimens obtained from mixed-oxide fuel pins irradiated in the Prototype Fast Reactor (PFR) to burnups of 4 and 9 atom percent. In these tests, specimens of fuel cladding were pressurized and heated until failure occurred.Samples of cladding from PFR fuel pins exhibited generally greater strength and ductility than specimens from Experimental Breeder Reactor-II (EBR-II) mixed-oxide fuel pins tested under similar conditions. Apparently, the PFR cladding properties were not degraded by a fuel adjacency effect (FAE) observed in fuel pin cladding from EBR-II irradiations.A recently developed model of grain boundary cavity growth was used to predict the results of the tests conducted on PFR cladding. It was found that the predicted failure temperatures for the relevant internal pressures were in good agreement with experimental failure temperatures.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

Online integrated visual inspection and sorting system for fast reactor fuels

Mixed oxide (MOX) fuel for prototype fast breeder reactor (PFBR) is designed to have initial burn up of 100,000 MWD/T. The major differences from thermal reactor fuel are relatively smaller dimension with central hole and higher plutonium concentration (21% and 28% of PuO₂) MOX pellets which are loaded into 2.5 m long clad tubes with depleted UO₂ blanket pellets at either end of the MOX stack. The relatively smaller dimension of fuel pellets for PFBR results in large volume at fabrication and inspection. To ensure fast and accurate inspection and sorting of as sintered pellets with less radiation exposure to personnel an integrated on line pellet inspection system for remote visual inspection and sorting of pellets based on diameter has been developed. Details of the integrated pellet inspection system developed at Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur along with the results of the performance trials has been described in this paper.     

Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

To analyze the effect of an inhomogeneous mixture of an PuO₂ powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO₂ fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known.     

A perspective on fast reactor fuel cycle in India

The growing energy needs of India can be fulfilled only by judicious mix of all the fuel resources. It is possible to achieve energy security and sustainability through the introduction of fast reactors in an expeditious manner and closing the fuel cycle. This approach is inevitable in view of the limited uranium resources in India. The Fast Breeder Test Reactor (FBTR) built by India uses mixed carbide as fuel and the 500 MW(e) Fast Breeder Reactor Project (PFBR), to be operational in 2010, will use mixed oxide as fuel. It has also been decided that fast reactors beyond 2020, with enhanced safety features and having better economy, will use metallic fuel. Having successfully operated FBTR with carbide fuels, we need to develop the fuel cycles for both the mixed oxide fuel in the near future and the metallic fuel expeditiously. The progress achieved so far and the plans for implementation are discussed in this paper.     

Neutron Behavior in Cluster-Type Fuel Lattices, (I)

Lattice parameters φ²⁸, φρ²⁵, ρ²⁸ and C^* were measured on cluster-type fuel lattices of the ATR (Advanced Thermal Reactor) by using a two region critical facility (D₂0-cluster test and H₂O-rod driver regions). Their dependence on lattice pitch, coolant-void ratio and fuel composition (whether UO₂ or PuO₂-UO₂) have been made clear by this experiment.

A foil handling technique has been developed for determining the lattice parameters of the Pu0₂- UO₂ fuel pins, and the resulting measurement errors are almost as small as those obtained on the U0₂ fuel pins.

The effects of the Cd-filter and of the presence of UO₂ buttons in the measurement of ρ²⁸ and χ²⁵ were studied experimentally and correction factors have been determined.

A method of observing the spatial distribution of the γ-ray source in an activated foil has been developed, and the relation between the spatial distribution and the coincidence counting efficiency of the foil has been examined.