Three-dimensional Neutron Kinetics Model Research Articles

Verification of three-dimensional neutron kinetics model of TRAP-KS code regarding reactivity variations

Abstract This work deals with TRAP-KS code verification. TRAP-KS is used for coupled neutron and thermo-hydraulic process calculations of VVER reactors. The three-dimensional neutron kinetics model enables consideration of space effects, which are produced by energy field and feedback parameters variations. This feature has to be investigated especially for asymmetrical multiplying variations of core properties, power fluctuations and strong local perturbation insertion. The presented work consists of three test definitions. First, an asymmetrical control rod (CR) ejection during power operation is defined. This process leads to fast reactivity insertion with short-time power spike. As second task xenon oscillations are considered. Here, small negative reactivity insertion leads to power decreasing and induces space oscillations of xenon concentration. In the late phase, these oscillations are suppressed by external actions. As last test, an international code comparison for a hypothetical main steam line break (V1000CT-2, task 2) was performed. This scenario is interesting for asymmetrical positive reactivity insertion by decreasing coolant temperature in the affected loop.

Space–time effects in the initiating phase of sodium fast reactors and their evaluation using a three-dimensional neutron kinetics model

The treatment of low probability events leading to core disruption is one of the key issues of R&D plans for the advanced reactor systems in general, and for sodium fast reactors in particular. Regarding sodium fast reactor systems, a major concern is that the core, at nominal, is not necessarily in its neutronically most reactive state. The simulation of the initiating phase of such accidents is of particular interest both for the prevention and the mitigation of routes leading to a large core disruption and recriticalities. Current analysis of the initiating phase relies on a Point Kinetics model, which neglects spatial variations of the neutron flux (i.e. the flux shape). Because the core geometry may be modified during such accidental conditions, the constant-shape approximation becomes questionable when applied to initiating phase simulations. The present paper investigates space–time effects during the degradation of a sodium fast reactor core. To this aim, a three-dimensional neutron kinetics model has been coupled to a multiple-channel thermal–hydraulics model and applied to evaluate spatial effects during a severe accident transient. As expected, it was found that spatial effects are not likely to be of importance in the pre-boiling portion of the transient reinforcing the use of a Point Kinetics model for the simulation of design-basis transient up to the boiling point. Departure from the Point Kinetics approach is noticeable after the fuel-pin breakup. The fuel being the source of neutrons, its motion within the core boundaries tends to create important spatial changes in the fission source. From this event, the Point Kinetics model fails in reproducing results obtained by the three-dimensional model and the study conducted in the present paper assesses this discrepancy.

Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests

TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.Consequently, (1)the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.

Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests.

TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1. Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version. Consequently, (1)the space-dependent power flow transitions in a...

TRACG Transient Analysis Code—Three-Dimensional Kinetics Model Implementation and Applicability for Space-Dependent Analysis

Space- and time-dependent phenomena, mostly related to neutron flux oscillations, have been observed in several boiling water reactor plants. A time-dependent three-dimensional transient analysis code is indispensable for simulating such phenomena. In a joint effort between the General Electric Company and the Toshiba Corporation, a three-dimensional neutron kinetics model has been implemented into the best-estimate thermal-hydraulics code, TRACG. A neutronics model implementation and the applicability of the modified TRACG code for analyzing space-dependent phenomena are discussed. To verify the code, startup tests with selected rod insertions, where control rods are locally inserted, are simulated. Both corewide, spatially in-phase neutron flux oscillations and regional, spatially out-of-phase oscillations are modeled. The results show that the modified TRACG code has sufficient capability to simulate space-dependent transients and is also a useful tool for investigating the fundamental mechanisms behind such transients.

On-Line Critical Control Rod Pattern Prediction Algorithm for BWR Plant Startup.

This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement.