Calculation of temperature coefficients of reactivity for EBR-II kinetic analyses

Component and regional temperature coefficients of reactivity for four loading configurations of the Experimental Breeder Reactor-II (EBR-II) are compared. The coefficients are calculated by summations of microcoefficients obtained by fine axial delineations of every subassembly. A special-sum method for obtaining effective coefficients for use in kinetics code channels representing subassembly groupings is described. Evaluations of rod-bank suspension coefficients and of grid-plate radial-expansion coefficients are also presented.     

Ways of altering the coefficients of reactivity in RBMK reactors

Conclusions In RBMK reactors there are many possibilities of acting on the coefficients of reactivity, primarily on the steam-void coefficient . Some of these methods can be implemented only by building new reactors and are irreversible (e.g., changing the lattice pitch of the fuel assemblies). Other ways of great interest make it possible to operatively act on the coefficients of reactivity even in existing reactors. These include such strong, but economically acceptable, measures as keeping some auxiliary absorbers in the reactor core or increasing the operational reactivity margin as well as economically effective measures involving an increase in the density of the fuel and in the initial enrichment. An extremely great effect on the steam-void coefficient of reactivity is displayed by such operating modes as maintaining the mean water density in the reactor and the energy distribution over the height at a required level.Translated from Atomnaya Énergiya, Vol. 46, No. 6, pp. 386–389, June, 1979.     

Evaluation of coupling coefficients in nodal/modal analysis

A procedure for developing neutron coupling coefficients for power-distribution calculations in LWRs using nodal analysis is described. The coefficients account for the fast and thermal flux variation within each node or fuel assembly by utilizing a 1-D diffusion-theory analysis of the flux ‘modes’ within the nodal volume. The procedure utilizes the fact that the fuel assemblies in most LWRs are arranged in such a manner that the reactor volume consists of a large number of neutronically self-sustaining zones. The paper provides both a physical and an analytical justification for the evaluation of neutron coupling coefficients in two energy groups.     

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF₂, LiF, ZrF₄ and Li₂BeF₄ eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Prompt and power reactivity coefficients for the next generation VVER-1000 reactor including hexagonal assemblies and annular fuels

Many neutronics as well as thermal-hydraulics calculations have been made to find the performance of the proposed annular fuels (internally and externally cooled fuel pins) for both next generation PWRs and BWRs. Specifically, there has not been a significant study on the Russian type VVER-1000 reactors with annular fuels. Our aim herein is to study two important safety coefficients of the Iranian VVER-1000 core including hexagonal annular fuel assemblies at its BOC. The safety coefficients are “prompt reactivity coefficient” and “power reactivity coefficient”, where all simulations are made using MCNP-5 code. We found less (absolutely) Doppler coefficient for the next generation VVER-1000 and therefore Doppler coefficient decreasing is a good feature to avoid more resonance neutrons absorbing in the U-238; causes more fission density and also less soluble boron for core controlling (at the BOC) with comparing to the current VVER-1000 solid pins.     

Core concept of a passive-safety fast reactor “METAL-KAMADO” and reactivity coefficients

New concept of a passive-safety simple fast reactor “METAL-KAMADO” with metallic fuels is presented, which has same concept as a passive-safety thermal reactor “KAMADO”. A fuel element of the “METAL-KAMADO” consists of metallic fuel (U–10%Zr) and cooling holes of He gas flow. These fuel elements are located in a reactor water pool of atmospheric pressure (0.1 MPa) and low temperature (<60 °C). In case of LOF, decay heats of fuel elements are removed by natural heat transfer from surfaces of the fuel elements to the reactor water pool.

Preliminary neutronic calculations of the “METAL-KAMADO” show possibility of high burn-up of more than 120 GWd/t with 10% enriched U–Zr fuel. Reactivity coefficients of the core are also discussed.     

Prospects of using different clad materials in a material test research reactor – Part 2 – The reactivity feedback coefficients

The reactivity feedback coefficients of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced with stainless steel-316 and zircaloy-4. Calculations were carried out to find the fuel temperature reactivity feedback coefficient, clad temperature reactivity feedback coefficient, moderator temperature reactivity feedback coefficient and moderator density reactivity feedback coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 38 °C to 50 °C, at the beginning of life, were maximum in magnitude for stainless steel-316 cladded fuel, followed by aluminum and least for the zircaloy-4 cladded fuel. The fuel temperature feedback coefficient increased in magnitude by 47.37% for stainless steel-316 and decreased by 4.72% for zircaloy-4 clad. The moderator temperature feedback coefficient increased in magnitude by 60.41% for stainless steel-316 and decreased by 3.03% for zircaloy-4 clad, while the moderator density feedback coefficient showed an increase in magnitude of 59.18% for stainless steel-316 and a decrease of 7.63% for zircaloy-4 clad. Zircaloy-4 gave a positive value for clad temperature feedback coefficient, while the others two did not have any clad temperature feedback coefficient.     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

ZGS Feedback System

In this paper, we investigate the beam phase damping techniques employed in the ZGS. Phase oscillation frequencies have been measured at various times during acceleration. Response data taken on all systems involved with phase and radial position measurement have been used to calculate the phase and radial position responses. These results compare favorably with responses measured on the ZGS with accelerated beam.