Analysis of Shutdown System Effectiveness in the Canadian Super Critical Water Reactor Using Coupled Thermal Hydraulics and Three-Dimensional Neutron Kinetics

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
David William Hummel ◽  
David Raymond Novog
Keyword(s):  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Flow Tube ◽  
Reactivity Effect ◽  
Neutron Kinetics ◽  
Power Pulse ◽  
Density Reduction ◽  
Positive Reactivity ◽  
Fuel Elements ◽  
Supercritical Water Cooled Reactor

Abstract The Canadian supercritical water-cooled reactor concept features a re-entrant fuel channel wherein coolant first travels down a center flow tube and then up around the fuel elements. Previous work demonstrated that in cases of sudden coolant flow reduction or reversal (such as that which would result from a large pipe break near the core inlet), the coolant density reduction around the fuel has a positive reactivity effect that results in a power excursion. Such a transient is inherently self-terminating since the inevitable density reduction in the center flow tube has a very large negative reactivity effect. Nevertheless, a brief power pulse would ensue. In this work, the possibility of mitigating the power pulse with a fast-acting shutdown system was explored. The shutdown system model, consisting of bottom-inserted neutron absorbing blades and realistic estimates of insertion rates and trip conditions, was added to a full-core coupled spatial neutron kinetics and thermal-hydraulics model. It was demonstrated that such a system can effectively mitigate both the peak magnitude of the power excursion and its duration.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis

2008 ◽  
Vol 238 (4) ◽  
pp. 1002-1025 ◽  
Author(s):  
F. D’Auria ◽  
S. Soloviev ◽  
V. Malofeev ◽  
K. Ivanov ◽  
C. Parisi
Keyword(s):  
Three Dimensional ◽  
Accident Analysis ◽  
Thermal Hydraulics ◽  
Neutron Kinetics

Validation of TRACG05 for Application to BWR AOO

2012 ◽  
Author(s):  
Asuka Matsui ◽  
Masashi Tamitani ◽  
Yoshiro Kudo ◽  
Sho Takano ◽  
Tatsuya Iwamoto ◽  
...  
Keyword(s):  
Experimental Data ◽  
Three Dimensional ◽  
Reactor Core ◽  
Expansion Method ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Neutron Kinetics ◽  
Safety Issues ◽  
Best Estimate ◽  
Assessment Matrix

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Kerntechnik ◽  
2011 ◽  
Vol 76 (3) ◽  
pp. 166-173
Author(s):  
U. Rohde ◽  
S. Baier ◽  
S. Duerigen ◽  
E. Fridman ◽  
S. Kliem ◽  
...  
Keyword(s):  
Block Type ◽  
Thermal Hydraulics ◽  
Neutron Kinetics

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

Kerntechnik ◽  
2016 ◽  
Vol 81 (4) ◽  
pp. 394-399 ◽  
Author(s):  
M. A. Uvakin ◽  
G. V. Alekhin ◽  
M. A. Bykov ◽  
S. I. Zaitsev
Keyword(s):  
Three Dimensional ◽  
Kinetics Model ◽  
Neutron Kinetics

scholarly journals Performance and applicability of a 2.5-D ice-flow model in the vicinity of a dome

2016 ◽  
Vol 9 (7) ◽  
pp. 2301-2313 ◽  
Author(s):  
Olivier Passalacqua ◽  
Olivier Gagliardini ◽  
Frédéric Parrenin ◽  
Joe Todd ◽  
Fabien Gillet-Chaulet ◽  
...  
Keyword(s):  
Flow Model ◽  
Three Dimensional ◽  
Thermal Conditions ◽  
Observation Data ◽  
Flow Tube ◽  
Ice Flow ◽  
Linear Temperature ◽  
Ice Surface ◽  
Scanning Window ◽  
D Model

Abstract. Three-dimensional ice flow modelling requires a large number of computing resources and observation data, such that 2-D simulations are often preferable. However, when there is significant lateral divergence, this must be accounted for (2.5-D models), and a flow tube is considered (volume between two horizontal flowlines). In the absence of velocity observations, this flow tube can be derived assuming that the flowlines follow the steepest slope of the surface, under a few flow assumptions. This method typically consists of scanning a digital elevation model (DEM) with a moving window and computing the curvature at the centre of this window. The ability of the 2.5-D models to account properly for a 3-D state of strain and stress has not clearly been established, nor their sensitivity to the size of the scanning window and to the geometry of the ice surface, for example in the cases of sharp ridges. Here, we study the applicability of a 2.5-D ice flow model around a dome, typical of the East Antarctic plateau conditions. A twin experiment is carried out, comparing 3-D and 2.5-D computed velocities, on three dome geometries, for several scanning windows and thermal conditions. The chosen scanning window used to evaluate the ice surface curvature should be comparable to the typical radius of this curvature. For isothermal ice, the error made by the 2.5-D model is in the range 0–10 % for weakly diverging flows, but is 2 or 3 times higher for highly diverging flows and could lead to a non-physical ice surface at the dome. For non-isothermal ice, assuming a linear temperature profile, the presence of a sharp ridge makes the 2.5-D velocity field unrealistic. In such cases, the basal ice is warmer and more easily laterally strained than the upper one, the walls of the flow tube are not vertical, and the assumptions of the 2.5-D model are no longer valid.

scholarly journals Heterogeneous reaction of HO<sub>2</sub> with airborne TiO<sub>2</sub> particles and its implication for climate change mitigation strategies

2017 ◽  
Author(s):  
Daniel R. Moon ◽  
Giorgio S. Taverna ◽  
Clara Anduix-Canto ◽  
Trevor Ingham ◽  
Martyn P. Chipperfield ◽  
...  
Keyword(s):  
Transport Model ◽  
Heterogeneous Reaction ◽  
Three Dimensional ◽  
Mitigation Strategies ◽  
Heterogeneous Reactions ◽  
Chemical Transport Model ◽  
Flow Tube ◽  
Tio2 Particles ◽  
The Impact ◽  
Change Mitigation

Abstract. One geoengineering mitigation strategy for global temperature rises resulting from the increased concentrations of greenhouse gases is to inject particles into the stratosphere to scatter solar radiation back to space, with TiO2 particles emerging as a possible candidate. Uptake coefficients of HO2, γ(HO2), onto sub-micrometre TiO2 particles were measured at room temperature and different relative humidities (RH) using an atmospheric pressure aerosol flow tube coupled to a sensitive HO2 detector. Values of γ(HO2) increased from 0.021 ± 0.001 to 0.036 ± 0.007 as the RH was increased from 11 % to 66 %, and the increase in γ(HO2) correlated with the number of monolayers of water surrounding the TiO2 particles. The impact of the uptake of HO2 onto TiO2 particles on stratospheric concentrations of HO2 and O3 was simulated using the TOMCAT three-dimensional chemical transport model. The model showed that by injecting the amount of TiO2 required to achieve the same cooling effect as the Mt. Pinatubo eruption, heterogeneous reactions between HO2 and TiO2 would have a negligible effect on stratospheric concentrations of HO2 and O3.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

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