Numerical modelling of modular high-temperature gas-cooled reactors with thorium fuel

The volumetric homogenization method for the simplified modelling of modular high-temperature gas-cooled reactor core with thorium-uranium fuel is presented in the paper. The method significantly reduces the complexity of the 3D numerical model. Hence, the computation time associated with the time-consuming Monte Carlo modelling of neutron transport is considerably reduced. Example results comprise the time evolutions of the effective neutron multiplication factor and fissionable isotopes (233U, 235U, 239Pu, 241Pu) for a few configurations of the initial reactor core.

Analyzing the Neutron Parameters in the Accelerator Driven Subcritical Reactor Using the Mixture of Molten Pb-Bi as Both Target and Coolant

In this paper, the Accelerator Driven Subcritical Reactor (ADSR) was simulated based on the structure of the TRIGA-Mark II reactor by the MCNPX program. The proton beam interacts on the Pb-Bi molten target with various energy levels from 0.5 GeV to 2.0 GeV. The important neutron parameters to evaluate the operability of ADSR were calculated as: the neutron yields according to various thicknesses of the target and according to the energy of the incident proton beam; the effective neutron multiplication factor for various fuel mixtures, along with its stability for some fuel mixtures; the axial and radial distributions of the neutron flux along with the height and radius of the core. The obtained results had shown a good agreement in using Pb-Bi molten as the interaction target and coolant for ADSR.

On the determination of neutron multiplication by the Rossi-alpha method

Introduction. This work contains the results of determining the prompt neutron multiplication factor in the subcritical state of a one-core BFS facility, obtained by the neutron coincidence method, for which the influence of the error in the βeff in determining the multiplication factor turned out to be insignificant. The core of the facility consisted of rods filled with pellets of metallic depleted uranium, 37% enriched uranium dioxide and 95% enriched plutonium, sodium, stainless steel and Al2O3. Stainless steel served as a reflector. Methods. In contrast to the inverse kinetics equation solving (IKES) method, which is convenient for determining reactor subcritical states, the neutron coincidence method practically does not depend on the error in the value of the effective fraction of delayed neutrons βeff. If in the IKES method the reactivity value is obtained in fractions of βeff, i.e., from the measurement of delayed neutrons, the neutron coincidence method is based on the direct measurement of the value (1 – kσp)2, where is the effective multiplication factor by prompt neutrons. The total multiplication factor is defined as keff = kσp + βeff. If, for example, keff ≈ 0.9 (which is typical for determining the fuel burnup campaign), then it is the error in determining kσp that is the main one in comparison with the error in βeff. Thus, a 10% error in βeff of 0.003–0.004 (typical for plutonium breeders) will make a contribution to the error 1 – keff equal to 1 – kσp + βeff ≈ 0.00035, i.e., approximately 0.35%, but not 10%, as in the IKES method. Rossi-alpha measurements were carried out using two 3He counters and a time analyzer. The measurement channel width Δt was 1.0 μs. From these measurements, the value of the prompt neutron multiplication factor was obtained. In this case, the space-isotope correlation factor for the medium with a source was calculated using the following values: Φ(x) – solutions of the inhomogeneous equation for the neutron flux and Φ+(x) – solutions of the ajoint inhomogeneous equation. Results. The authors also present a comparison of the results of the Rossi-alpha experiment and measurements of the BFS-73 subcritical facility by the standard IKES method in determining the multiplication factor value. The data of the IKES method differ insignificantly from the results of the Rossi-alpha method over the entire range of changes in the subcriticality with an increase in the subcriticality of the BFS-73 one-core facility. Conclusion. It was impossible to apply the neutron coincidence method to fast reactors; however, the method turned out to be quite workable on their models created at the BFS facility, which was successfully demonstrated in this study.

INTRODUCING BURNUP CREDIT FOR EXPANSION STAGE 2 OF THE EXTERNAL SPENT FUEL POOL AT GÖSGEN NPP

We present a summary of the actinide-plus-fission-product burnup credit criticality safety licensing analysis for Expansion Stage 2 (ES2) of the external spent fuel pool at Gösgen nuclear power plant. In ES2, the nine Expansion Stage 1 storage racks currently installed in the external spent fuel pool are going to be supplemented by nine ES2 storage racks with a significantly reduced fuel assembly pitch. They are designed for loadings with fuel assemblies above a well-defined minimum required burnup. The objective of the criticality safety analysis is to calculate the minimum required burnup values for the uranium and MOX fuel assemblies to be stored in the ES2 storage racks. We use a methodology that allows us to take into account the reactivity effects due to variabilities and uncertainties of all relevant parameters involved in a burnup credit criticality safety assessment in a bounding manner. These include manufacturing tolerances of the fuel assemblies and storage racks, the irradiation histories and burnup profiles of the spent fuel assemblies, the bias of the depletion code used to calculate the isotopic inventories of the irradiated fuel, and the bias of the criticality code used to calculate the neutron multiplication factor of the considered storage configuration. A combination of different statistical procedures is used to evaluate and propagate the uncertainty information on the input parameters and translate it into statistical confidence statements about the neutron multiplication factor. It should be noted that the presented analysis is related to the first implementation of a significant burnup credit for wet storage of PWR fuel in Switzerland.

SUBCRITICALITY ESTIMATION BY VIRTUAL NEUTRON CAPTURE METHOD

The criticality safety control technique is required for the fuel debris removal from the Fukushima Dai-Ichi Nuclear Power Station which experienced a severe accident. The subcriticality estimation is expected to be done with only limited information about the fuel debris while the primary containment vessel internal survey work is ongoing. The purpose of this study is to develop the subcriticality estimation method called the virtual neutron capture method. The neutrons from the surface of the fuel debris represent a major portion of detector counts. The method consists of two evaluations: the evaluation at the surface of the fuel debris for which the isotope compositions are known by fuel debris sampling and the evaluation at the region of the fuel debris for which these compositions are unknown. For the unknown composition region, the average isotope composition with arbitrary water content is given. The method surveys the relationship with the detector count and the neutron multiplication factor with any size of the unknown composition region and any ratio of the water content before the on-site evaluation. The method is verified by experiments done in the Kyoto University Critical Assembly. The method shows that the maximum difference from the reference neutron multiplication factor is 4.5 %dk. As a result, the virtual neutron capture method can be adopted to the subcriticality monitoring if the method includes the estimation margin of 4.5 %dk within the neutron multiplication factor range from 0.70 to 0.95.