Spatial corrections for reactivity measurement in lead accelerator driven system

In the subcritical nuclear reactor, the subcriticality is closely related with the neutron flux distribution. Since the relationship between them is complex, the subcriticality measurement by conventional methods involves some difficulties, particularly when the subcriticality is large. A new subcriticality measurement method is proposed for the steady subcritical state. By this method, the relative distribution of the change in neutron flux between before and after the perturbation is numerically calculated. On the other hand, the amplitude of the perturbation is estimated from the count rates of neutron detectors. By this combination of the calculation and experiment, the subcriticality can be accurately estimated, to the extent that the perturbation is qualitatively known beforehand. One of the advantages of this method over the conventional modified neutron source multiplication method is that numerical calculations are only executed on unperturbed states. In addition, the completeness of the eigenfunctions is not assumed. The validity of this method is confirmed through the numerical calculation.

Building a dynamic code to simulate new reactor concepts

System Analysis Module (SAM) is being developed at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report presents the development of Point-Kinetics modeling and reactivity feedback mechanisms in SAM. It also presents a new capability for predicting the thermal expansion in various structural components by coupling SAM with an external thermomechanics module (Tensor Mechanics) from the MOOSE framework. The analysis of the transient behavior of a nuclear reactor requires the coupled simulation of reactor kinetics and thermal-hydraulics of the reactor core, especially for those unprotected transients where the reactor scram system may not function properly. The point kinetics model has been widely used for reactor safety analysis due to its simplicity to capture the transient behavior of the reactor. Various reactivity feedback models have been developed and integrated with the Point-Kinetics module, including fuel axial expansion, core radial expansion, fuel Doppler, and coolant density reactivity. The reactivity feedback models in SAM are similar to the respective models used in SAS4A/SASSYS-1. This report first presents the brief theory of the Point-Kinetics module and reactivity feedback models. A number of verification tests have been performed where the code simulations are compared to the analytical model results. The reactivity feedback due to the thermal deformation, such as the fuel axial expansion and core radial expansion, is important for SFR transient analysis. Simplified thermal expansion models for the fuel pin and reactor constrain system (e.g. grid plate) are developed and verified in SAM. Additionally, a coupling interface is developed to couple SAM with external thermomechanical analysis modules for more accurate predictions of the thermal expansion of different components during the transients. The current coupling interface has been tested with the Tensor Mechanics module from MOOSE. These point kinetics and reactivity feedback modeling capabilities have also been demonstrated by simulating the early stage of the unprotected loss-of-flow (ULOF) accident in the Advanced Burner Test Reactor (ABTR). Both the stand-alone SAM and coupled SAM and Tensor Mechanics simulations are performed. It is confirmed that the major physics phenomena in the heat transport system of the ABTR reactor are captured by SAM, and the point kinetics model, reactivity feedback models, and the coupling schemes are working well as expected.

Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization

Long-lived fission products and minor actinides produced in nuclear power plants are the most radiotoxic nuclear wastes. They can be transmuted into stable nuclei or into nuclei with shorter lifetime thanks to the so-called Accelerator Driven Systems (ADS), consisting of the coupling of an intense high energy proton beam, a spallation target and a sub-critical reactor core. For safety reasons, an on-line and robust measurement of the reactivity during loading and power operation is mandatory. The investigation of the relationship between the current of the accelerator and the power level (or neutron flux) of the reactor appears to be powerful, any change in reactivity being accessible through the measurement of the current and the flux. Such a relationship will be studied in an experiment to be performed at the YALINA facility (JIPNR Sosny - Belarus) in the framework of the EUROTRANS IP (6 th FP). At this installation, 14 MeV neutrons are produced in T(d,n) 4 He reactions by a deuteron beam impinging on a TiT target. Due to the tritium consumption over time, the intensity of the deuteron beam cannot be used for the monitoring of the neutron beam. The source neutron yield itself has to be accessed. This contribution describes the performance of a three-element silicon telescope dedicated to the detection of recoil protons produced in reactions induced by 14 MeV neutrons.

At the University of Cincinnati Nuclear Engineering Program, the resurrection of a once moth-balled subcritical reactor facility now provides an ideal laboratory experience to its students. Its simplicity and accessibility are among its key features. This article focuses upon supplementing the experimental aspects of this facility via the development of a three-dimensional, full detail, MCNP model of this reactor, while emphasizing the validation of this computational tool against laboratory measurements. The subcritical reactor parameters herein compared include: neutron and photon flux distributions, fission rate, and the subcritical multiplication factor (k-effective), among other relevant parameters of interest.

The point kinetics model for an initially subcritical system such as the accelerator-driven system (ADS) requires the adjoint-weighted kinetics parameters to be estimated in its reference subcritical state with external sources. Monte Carlo (MC) algorithms are newly developed to calculate the kinetics parameters weighted by the adjoint function, which is the solution to the inhomogeneous adjoint equation, in the MC fixed-source mode calculations. The developed algorithm is verified through a simple two-group problem by comparing with analytic solutions. The one-point kinetics analyses using the estimated kinetics parameters are applied for beam trip transient problems and a control rod drop experiment in ADS at the Kyoto University Critical Assembly.

In this paper, the solution of the Neutron Point Kinetics model is presented, adding the effects of temperature and absorbers poisons within a historical and technical context to simulate the preliminary characteristics of the Chernobyl accident. The Point Kinetics model was able to extract physical information consistent with what was expected to predict the reactor situation until the accident. It was also possible to verify, given the results, that the Rosenbrock method was able to overcome the degree of stiffness of the ODE system, besides solving a non-linear problem. Thus, this study has contributed to highlighting the importance of temperature effects and especially absorbers poisons in the final power behavior, extremely relevant for decision making in the operation and safety of a nuclear power plant.

A subcritical nuclear reactor, Model 9000, Nuclear Chicago, is installed and operating at the Aristotle University of Thessaloniki, in the Atomic and Nuclear Physics Laboratory, at Thessaloniki, Northern Greece. The fuel is about 5500 lbs (2495 kgs) natural uranium metal (U3U8), the moderator about 3600 lbs (1633 kgs) light water, H2O and the reflector is also light, water, H2O. The lattice core is hexagonal, 42 inches (1.07 m) high and of 35 inches (88.90 cm) maximum diameter. The neutron source at the core is Am-Be 5 curies (185 GBq), 1.1·107n·s−1. The reactor is used for the activation of various materials by neutrons such as indium, the determination of the thermal neutrons flux, the horizontal and the vertical distribution of the neutron flux, material buckling, B, and geometric buckling, B, the parameters of the reactor, and the albedo of water for thermal neutrons with foils of indium.

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors and accelerator-driven systems (ADS). The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and ADS makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a neutron reflector opens a possibility for substantial elongation of prompt neutron lifetime. As a result, chain fission reaction in the reactor core could be slowed down, and the reactor operation could become safer. 5. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket. Thus, favorable conditions could be formed in the ADS blanket for effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

Experimental studies on the thorium-loaded accelerator-driven system (ADS) have been conducted at the Kyoto University Critical Assembly. Mockup experiments are carried out in both the critical and subcritical states to investigate the profile of the neutron flux for thorium capture and fission reactions. Thorium plate irradiation experiments for thorium capture and fission reactions demonstrate fission reactions in the critical state, and the calculated versus experimental C/E values of the reaction rates show accuracy within a relative difference of about 30 %. In the ADS experiments with an external neutron source (14 MeV neutrons and 100 MeV protons), subcritical experiments were carried out in the thorium-loaded cores to investigate the profile of neutron flux for capture reactions by the measurement of 115In(n, γ)116mIn reactions. The results reveal the difference between reaction rate distributions attributed to varying not only the neutron spectrum of the core but also the external neutron source at the target location. A comparison between the measured and calculated reaction rate distributions reflects the accuracy of reaction rate analyses for the thorium-loaded ADS experiments with an external neutron source. Additionally, kinetic experiments were carried out to deduce the delayed neutron decay constants and subcriticality by the pulsed-neutron method.

The paper presents the design objectives and structure of China’s accelerator-driven subcritical system subcritical assembly Venus-1. The core of Venus-1 is a coupled core with a fast neutron zone and a thermal neutron zone. The fast neutron zone is at the center of the core and is formed by natural uranium fuel. A fast neutron spectrum field can be produced in the fast neutron zone and used for transmutation. The thermal neutron zone surrounds the fast neutron zone and is formed by low-enriched uranium fuel; it is a fission zone. Venus-1 is driven by an Am-Be or other steady neutron source (252Cf, D-D reaction or D-T reaction) to research the effect of an external neutron source with different energies on it or is driven by a D-T pulsed neutron source through the China Institute of Atomic Energy pulsed neutron generator to research the dynamic characteristic. On July 18, 2005, the first fuel element was loaded into the Venus-1 subcritical assembly, and some preliminary experiments on subcritical and neutron distribution have been performed. The final load has been determined by preliminary experiments on Venus-1. The relative neutron importance of the external source, the relative distribution of the thermal neutron flux in the fast neutron zone, and the variation of the neutron count with the addition of natural uranium fuel also have been determined by preliminary experiments on Venus-1.

Experimental analysis for neutron multiplication by using reaction rate distribution in accelerator-driven system

A new technique has been developed for monitoring slow reactivity changes in Canadian deuterium uranium (CANDU) nuclear reactors during low power operation, following a sustained period of high power operation. The power doubling halving worth (PDHW) test tracks slow reactivity changes by evaluating the reactivity perturbation required by the power regulating system to halve and double reactor power over time. During low power operation of a CANDU reactor, the PDHW test is used to monitor the decay of the photoneutron precursors, so that the reactor power can be lowered using a preset amount of reactivity. The PDHW test is described in this paper and is validated by using computer simulations and operating data from a CANDU reactor.

A non-intercepting monitoring method for beam position on the target in an accelerator driven system