Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm³ by the fuel of density 4.00 gU/cm³ then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm³ in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29–62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.     

Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

Resonance Scattering Near the Thermal Region

In this paper the effect on the neutron flux, in a homogeneous mixture of moderator and fuel, of scattering from low energy resonances in the fuel is considered. Particular attention is given to resonances for which slowing down theory is inappropriate. The kernel for a monatomic gaseous resonance scatterer is derived and compared with the ‘slowing down’ kernel commonly used. The effects of the two kernels on the flux near the resonance for the particular case of ²⁴⁰Pu and a light moderator are also compared.     

An Analysis of Continuous Bi-Directional Reactor Refueling

An analysis of continuous bi-directional reactor refueling is developed by one group diffusion approximation. The fundamental integro—differential equation is first converted to second order nonlinear differential equation, which is further reconverted to first order nonlinear differential equation, to finally take the form (du/dx) ²=f(u). By applying the elliptic function theory, the analytical solution for neutron flux and eigen value is obtained, and their physical characteristics are examined. Numerical results for neutron flux distribution are presented for the case where the rate of fuel feed movement is not the same for the two directions.     

Neutronic and thermal hydraulic design of the graphite moderated helium-cooled high flux reactor

A new design concept for a high flux reactor was investigated, where a graphite moderated helium-cooled reactor with pebble fuel elements containing (²³⁵U, ²³⁸U)O₂ TRISO coated particles was taken as its design base. The reactor consists of an annular pebble bed core, a central heavy water region, and inner, outer, top, and bottom graphite reflectors. The maximum thermal neutron flux in its central heavy water region as high as 10¹⁵ cm⁻² s⁻¹ with thermal neutron spectral purity of more than two orders of magnitude and a large usable volume of more than 1,000 liters can be achieved by (1) diluting the fissile material in the core and (2) optimizing the core-reflector configuration. The in-core thermal-hydraulic analysis was done to obtain adequate information about the coolant flow pattern and pressure distribution, core temperature profile, as well as other safety aspects of the design. To protect the reactor during off-normal or accident events, a large margin of temperature difference between the operating fuel temperature and the fuel limit temperature is confirmed by lowering the coolant inlet and core rise temperatures.     

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

A multigroup method of calculation is presented for describing the neutron behavior in a cluster-type fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).

The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO₂ or PuO²+UO² fuel pins, with heavy-or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

Recoil Release of Fission-Gas from UO2

The fission-gas released from in-core UO₂ pellets was investigated with the helium-swept fuel irradiation facility installed in the Hitachi Training Reactor. It was found that the release rate of fission-gas increased gradually as irradiation continued for 5hr under a constant thermal neutron flux of 1.0×10¹² n/cm²·sec and at a fuel temperature of about 150°C. This observation indicates that a recoil-activated release mechanism may be more effective than direct-recoil release in the case of clad fuel in the region of temperature where fission-gas release rate is temperature independent.     

Preliminary study of high neutron flux fusion heating

A heating scheme for nuclear fusion is proposed based on the availability of a high flux, low energy neutron source. The heat is derived in the reaction ⁶Li (n, T) ⁴He resulting from the incidence of a low energy neutron beam on a sample of ⁶Li D. The energy release per reaction, Q = 4.6 MeV, is converted through electron Coulomb collisions thereby quickly dissociating the solid sample to the plasma state. For 10⁻³ eV neutrons it is estimated that this dissociation occurs in 7 ms for an incident flux of 10¹⁷ cm⁻² · s⁻¹. The possibility of further driving the heated fuel to fusion is also discussed.     

The materials test station: A fast-spectrum irradiation facility

The United States Department of Energy is developing technologies needed to reduce the quantity of high-level nuclear waste bound for deep geologic disposal. Central to this mission is the development of high burn-up fuel with significant inclusion of plutonium and minor actinides. Different fuel forms (e.g., nitrides, oxides, and metal matrix) and composition are under study. The success of these cannot be judged until they have been irradiated and tested in a prototypic fast neutron spectrum environment. In 2005, the US Congress authorized funding for the design of the materials test station (MTS) to perform candidate fuels and materials irradiations in a neutron spectrum similar to a fast reactor spectrum. The MTS will use a 1-MW proton beam to generate neutrons through spallation reactions. The peak neutron flux in the irradiation region will exceed 1.2 × 10¹⁹ n m⁻² s⁻¹ and the fast neutron fluence will reach 2 × 10²⁶ n m⁻² per year of operation. Site preparation and test station fabrication are expected to take four years.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Buckling Correction for Experimental Power Distribution within Fuel Assemblies in a Finite System

For design calculations to determine the local power distribution in a fuel assembly of a power reactor, the neutron flux is usually assumed to be symmetrical at the outer boundary of the assembly. In actuality, experimental data on power distributions are obtained in a finite system where this symmetry does not apply, so that the calculated values cannot be directly compared with observed data. In a zero power critical experiment in particular, the measurement must be performed in a fairly small core assembly so economize the amount of fuel materials to be used for simulation. This introduces the necessity of special considerations in the comparison between design and observed data.

The authors propose a method incorporating direct corrections to the experimentally determined. power distributions based on the geometrical buckling of the system. In this method the experimental power P ⁰ (r) is divided by the neutron flux Q (r) which is determined in the critical state with geometrical buckling in a bare (one neutron energy group) reactor, neglecting the reflector region of the experimental system. A sample application of the method to an actual light water lattice has confirmed the validity of the method.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Reactivity Control Capability of Fuel-Salt Processing System in a Molten-Salt Breeder Reactor

An evaluation is made of the reactivity control capability of the fuel processing system (FPS) in a molten-salt breeder reactor. The principal functions required of the FPS are : (a) Isolation of ²³³Pa from regions of high neutron flux during its decay to ²³³U, and (b) the removal of fission products from the system. The FPS can very usefully serve also to control the primary system reactivity by appropriately utilizing its function of extracting uranium and reconstituting the fuel contained in the salt. The principles of operation are quite similar to the chemical shim control system currently installed in PWR's whereby the core reactivity, affected by changes in the moderator temperature, fuel burnup and transient Xe, is adjusted by regulating the concentration of boric acid introducted in the moderator as neutron absorber. The present study examines the capability of the FPS to follow transient Xe as in PWR's, and proves that the FPS should effectively serve as a system for adjusting not only long-term changes in reactivity but also short-term transient variations without any accompanying difficulties foreseen in operation.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10³ ∼ 10⁷) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.     

Adaptive burnup stepsize selection using control theory for 2-D lattice depletion simulations

A control theory approach is adopted to determine the temporal discretization during two-dimensional lattice physics depletion simulations. Two primary applications of automated and adaptive stepsize control are identified: (i) the presence of strong absorbers such as gadolinium, where the accurate burnout of the isotopes requires a depletion stepsize smaller than typically required, and (ii) high fidelity multiphysics simulations, e.g. loosely coupled physics, where the coupled physics are nonlinear in time and stepsize changes may be necessary to obtain an accurate coupled solution. A conventional predictor–corrector method is used to address the nonlinearity of the nuclide transmutation and neutron flux. An adaptive stepsize method is developed based on monitoring the one-group scalar neutron flux at both the predictor and corrector steps to approximate the convergence residual of the nonlinear solution. A user-specified tolerance on the L² relative error norm of the scalar neutron flux is utilized by the stepsize controller. Controllers that include integral, proportional, and/or derivative components are investigated and parameterized using Latin hypercube sampling of the controller input parameters. Three distinct fuel loadings of pressurized water reactor 17 × 17 fuel pin assemblies are considered, including no burnable absorbers, Integral Fuel Burnable Absorber, and gadolinium fuel pins. The required depletion stepsizes, as predicted throughout the cycle by the controller, are compared with a very small stepsize (0.01 MW d/kgHM) reference solution and a solution obtained by a typical rule of thumb depletion stepsize sequence.     

In-pile dimensional changes in neutron irradiated zirconium base alloys

A model is developed to predict in-pile growth in zirconium base alloys as a function of neutron flux, neutron fluence, temperature, dislocation density, and texture. The model is based on vacancy and interstitial behavior with respect to straight dislocations, dislocation loops, depleted zones and grain or sub-grain boundaries. Results indicate very little growth dependence on temperature or neutron flux at temperatures below ~320°C, at fluxes above ~10¹³ n/cm² sec, and at fluences below 10²¹ n/cm². As the flux is lowered and the temperature and fluence are raised, the temperature and flux dependencies increase. Comparison between theory and data is given for both growth and dislocation loop size.