Thermal Reactor Core Research Articles

Simulation on the Effect of Coolant Inlet Temperature and Mass-Flowrate Variations to the Temperature Distribution in Single Pellet Thermal Reactor Core

An important factor in the development of nuclear energy is reactor safety. The performance of heat transfer from nuclear fuel to coolant is the main key to the reactor safety. This paper presents simulation on temperature distribution in two-dimensional laminar flow for single pellet thermal reactor with variation on temperature inlet and mass-flowrate. The OpenFoam platform (SimFlow 3.1) has been used for the computational and numerical analysis. The simulation is carried out on a single pellet with an aspect ratio of 1.2. The variations in the mass velocity of the coolant flow are 10, 100, and 14300 kg×s-1 with a constant coolant temperature of 552 K, and the variations of the input coolant temperature are 300, 552, and 1000 K with a constant mass-flowrate of 10 kg×s-1. The results obtained from the simulation show that for variations in the input coolant temperature of 300, 552, and 1000 K, the fuel temperature can be reduced respectively by 34, 26, and 14 K. At the fastest variation in the coolant mass-flowrate of 14300 kg×s-1, the coolant temperature around the pellet rises by 396 K. The decrease in fuel temperature is significant if the mass-flowrate of the input coolant flow is relatively low.

Transient analysis of a subcritical reactor core with a MOX-Fuel using the birth-and-death model

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction β for 239Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

On similarity of various reactor spectra and 235U prompt fission neutron spectrum

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 – 6 MeV. Above this energy the reactor spectra becomes identical with the 235U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra.

NEMHEX solution of benchmark problems on thermal and fast breeder reactor cores with hexagonal lattice

This paper gives the solution of some benchmark problems for reactor core calculations in hexagonal geometry using a computer code NEMHEX (Nodal Expansion Method for HEXagonal geometry). The problems cover a wide range of thermal reactor cores including large heavy water, fast breeder and VVER reactors in both two and three dimensional. The code is based on higher order nodal expansion method (NEM) for the solution of the neutron diffusion equation in 3-D hexagonal geometry. In this method, flux in each mesh/node and energy group is expanded in terms of polynomial up to third order in X–Y plane and fourth order in axial direction (Z). Weighted residual method is used to calculate the higher order expansion coefficients of the polynomials. Both the partially integrated radial and axial leakages are approximated by the quadratic polynomial. Higher order polynomial expansion [with respect to the work by Lawrence (DIF3D nodal neutronics option for two-and three-dimensional diffusion theory calculations in hexagonal geometry, 1983)] of flux as well as leakage in each node improves the accuracy in power distribution of VVER-1000 type reactor (which is having large size lattice) in terms of maximum reduction in error from 8 to 2%. The final inhomogeneous response matrix equation involves spatial moments of nodal flux and surface averaged partial currents across the surfaces of node. These equations are solved using conventional fission source iteration. This code is used for calculation of effective multiplication factor, neutron flux and fission power distribution. The results of the analysis establish the adequacy of NEMHEX for thermal as well as fast reactor core calculations in hexagonal geometry.

The Possibilities of Fission Material Reproduction Increase in Thermal Reactor with the Assemblies with a Hard Neutron Spectrum

This paper addresses the problem of fission material reproduction increase in thermal reactors. Reproduction increase is achieved while decreasing the fission material content in the fuel. In that case, a decrease of neutron loss in construction materials and a neutron leakage decrease are required for obtaining reactor criticality. Effectiveness of the reactor functioning can be increased by the use of additional neutron sources, for example, n‐2n reaction in beryllium framing. The possibility of additional neutron source development with the use of fast neutrons with an energy distribution close to the fission spectrum in the major part of thermal reactor core is researched in this paper.

Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors

A new adaptive stochastic approximation method for an efficient Monte Carlo calculation of steady-state conditions in thermal reactor cores is described. The core conditions that we consider are spatial distributions of power, neutron flux, coolant density, and strongly absorbing fission products like 135Xe. These distributions relate to each other; thus, the steady-state conditions are described by a system of nonlinear equations. When a Monte Carlo method is used to evaluate the power or neutron flux, then the task turns to a nonlinear stochastic root-finding problem that is usually solved in the iterative manner by stochastic optimization methods. One of those methods is stochastic approximation where efficiency depends on a sequence of stepsize and sample size parameters. The stepsize generation is often based on the well-known Robbins-Monro algorithm; however, the efficient generation of the sample size (number of neutrons simulated at each iteration step) was not published yet. The proposed method controls both the stepsize and the sample size in an efficient way; according to the results, the method reaches the highest possible convergence rate.

Three-Dimensional Benchmark Tests of ENDF/B-VI for Thermal Reactor Core

Critical experiments of several thermal reactor cores are analyzed with ENDF/B-VI to assess its applicability to light water reactor analyses. The benchmark tests chosen use the continuous energy Monte Carlo code MCNP with exact treatment of a three dimensional geometry, and a cross section library is prepared by using the NJOY91 code. Calculated keff's underestimate experimental values from 0.3 to 0.9%Δk, and differences between calculated and measured keff's linearly increase with 238U effective capture rate. With a view to investigating the differences, sensitivity analyses are performed by minor modification of the 238U capture cross section at resonance as well as thermal energy regions. As a result, calculated keff's agree well with experimental results by changing the 238U resonance parameters to those of Liou and Chrien below 120eV and decreasing the thermal capture cross section value by 2.3%. These modifications improve the keff's of the TRX-1 and TCA 3.00U cores by 0.8 and 0.3%Δk, respectively.

Three-Dimensional Benchmark Tests of ENDF/B-VI for Thermal Reactor Core.

Three-Dimensional Benchmark Tests of ENDF/B-VI for Thermal Reactor Core.

Determination of Effective Diffusion Parameters in Thermal Reactor Assemblies

Effective diffusion parameters have been determined so as to preserve reaction rates in a super-assembly consisting of few assemblies with different neutronic properties. The effective diffusion parameters are used for few-group diffusion calculations in one-and two-dimensional models of thermal reactor cores. The results are compared with those from reference heterogeneous calculations which take into account heterogeneity in each assembly within the framework of diffusion theory, and those obtained by using the conventional flux-weighted group constants. The use of the effective diffusion parameters results in a much more accurate prediction of the neutron multiplication factor and power densities than those predicted using the conventional flux-weighted group constants when using fine meshes.

Determination of effective diffusion parameters in thermal reactor assemblies.

Effective diffusion parameters have been determined so as to preserve reaction rates in a super-assembly consisting of few assemblies with different neutronic properties. The effective diffusion parameters are used for few-group diffusion calculations in one-and two-dimensional models of thermal reactor cores. The results are compared with those from reference heterogeneous calculations which take into account heterogeneity in each assembly within the framework of diffusion theory, and those obtained by using the conventional flux-weighted group constants. The use of the effective diffusion parameters results in a much more accurate prediction of the neutron multiplication factor and power densities than those predicted using the conventional flux-weighted group constants when using fine meshes.

Servo-Analysis and Stablization of Xenon Oscillations in Large Thermal Nuclear Reactor Power Plants

A two-point coupled core reactor kinetics model (Baldwin's model) is used to derive formulae which predict the period and damping factor of the xenon oscillations for a thermal reactor core in terms of the xenon and iodine kinetics parameters and the core parameters and coupling coefficient. The damping factor is a measure of the core stability with respect to the xenon oscillations. The dependence of the core coupling coefficient as a function of the square of the ratio of the core migration length to the core dimension is derived. Calculations are made using PWR core parameters which show that the stability with respect to the xenon oscillations is reduced by increased power density, increased core size, and decreased power reactivity feedback. The periods and damping factors predicted are within the range observed in PWR xenon oscillation experiments. A servo-analysis of the xenon oscillations is made using frequency response techniques. The stability margins are computed, and servo-analysis techniques are utilized in synthesizing a feedback control system to increase the xenon stability margins and effectively damp the xenon oscillations. A practical system for accomplishing this is proposed.

ALTEMP 601

Abstract ALTEMP 601 is a nickel-chromium solid-solution alloy with excellent high-temperature properties with good resistance to oxidizing, carburizing, and sulfur-containing environments. The properties make it attractive for applications in aerospace and process industries as well as automotive thermal reactor core components. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties as well as fracture toughness, creep, and fatigue. It also includes information on high temperature performance and corrosion resistance as well as forming, heat treating, machining, and joining. Filing Code: Ni-182. Producer or source: Allegheny Ludlum Corporation.

An Investigation of Fast-Reactor Fuel Testing in a Thermal Reactor Core

This paper investigates the feasibility of testing fast-reactor fuel elements in the core of a thermal reactor. A fast-reactor environment is approximated by placing an assembly of fast-reactor fue...