A Simplified Two-Group Multipoint Kinetics Model

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Santosh K. Pradhan ◽  
K. Obaidurrahman ◽  
Kannan N. Iyer
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled System ◽  
Kinetics Model ◽  
Primary Circuit ◽  
Thermal Hydraulics ◽  
Time Step ◽  
Large Power

Abstract Detailed multiphysics modeling of nuclear power plants has become a necessity in the era of best-estimate analysis. For a number of transients with strong coupling between the neutronics in the reactor core and the fluid-dynamics in the primary circuit and overall heat transfer, it is required to carry out coupled system thermal hydraulics and core three-dimensional (3D) neutronics analysis. Point kinetics approach in the system thermal-hydraulics (TH) code RELAP5 limits its use for many reactivity-induced transients, which involve asymmetric core behavior. In a recent development, a simplified multipoint kinetics model has been coupled with system TH code RELAP5 to circumvent its inadequacy for the analysis of reactivity-induced transients involving asymmetric core behavior. The objective of this paper is to validate the simplified multipoint kinetics model against an asymmetric fast transient benchmark problem in a large power reactor. Time-step and nodalization sensitivity studies have been performed. It is demonstrated that the multipoint kinetics model results are in good agreement with the benchmark, advocating its applicability.

Validation of a CFD Code for the Analysis of Hydrogen Behaviors and Thermal Hydraulics in Containments

2018 ◽  
Author(s):  
Meilan Chen ◽  
Zeming Zheng
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Convection Heat Transfer ◽  
Navier Stokes ◽  
Thermal Hydraulics ◽  
Navier Stokes Equations ◽  
Water Metal ◽  
Water Reactors

During the process of core melt-down accident in light water reactors, large quantities of hydrogen generated by drastic water-metal reaction are released to the containment. Subsequently, hydrogen-rich layer may be formed under the dome of the containment, threatening the integrity of nuclear Power Plants (NPPs). In the framework of a China national R&D project, China Nuclear Power Research Institute (CNPRI) has developed a three dimensional CFD Code for the assessment of hydrogen behaviors and relative thermal hydraulics in containment. The code solves the time-dependent Navier-Stokes Equations with multi-gas species. Validation with International Standard Problems (ISP) and other test data based on a Phenomena Identification and Ranking Table (PIRT) has been undergoing together with the development of this code. In this paper, the test cases of HYJET, COPAIN and TOSQAN 101 Test are validated. Stratification, buoyancy induced mixing in gases, convection heat transfer and condensation on surface are evaluated in the former two cases, while gas entrainment and mixing by spray droplets in the later one. Excellent agreements between experimental data and model predictions are obtained. In order to meet the requirements for application of the code in practical NPP design and safety analysis, further validations of other phenomena in PIRT should be performed in the near future.

Coupled Neutron Dynamic and Thermal Hydraulic Models of the HTR Simulator

2013 ◽  
Author(s):  
Zhe Sui ◽  
Jun Sun ◽  
Chunlin Wei ◽  
Yuanle Ma ◽  
Wenzhi Shan
Keyword(s):  
Nuclear Power ◽  
Dimensional Space ◽  
Three Dimensional ◽  
Reactor Core ◽  
Hydraulic Modeling ◽  
Thermal Hydraulics ◽  
Dynamic Simulations ◽  
Hydraulic Models ◽  
Coupling Techniques ◽  
Dimensional Space Time

Along with the design and construction of the high temperature reactor pebble bed module (HTR-PM), the engineering simulator system (ESS) has been accomplished to deal with the key techniques of the HTR simulator. In our previous papers, the three dimensional space-time neutron dynamics and the thermal hydraulic modeling of the HTR reactor core were introduced. In this paper, we concentrated on the detailed coupling techniques of the neutron dynamic and thermal hydraulic models, which was one of the most important assurances to the dynamic simulations of the HTR-PM. The mechanism of minus temperature feedback effect of the HTR was introduced by the materials participated in the coupling calculations, as well as those parameters being transferred. In parallel calculations, the neutron dynamics and thermal hydraulics were coupled by exchanging the power density and the temperature in fuels, moderates, reflectors and so on. With complete reactor core model, the coupling calculations of the neutron dynamics and thermal hydraulics were tested in many static and dynamic cases to show good performances. Based on that model ability, the ESS can simulate the start-up, shut-down and several accidents for the whole nuclear power plant.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

scholarly journals Assessment of water hammer effects on boiling water nuclear reactor core dynamics

2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah
Keyword(s):  
Pressure Wave ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Pressure Wave Amplitude ◽  
Sensitivity Studies ◽  
Rapid Transients ◽  
Complex Phenomena

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

An Efficient, Flexible Arrangement to Generate High Quality Process or Domestic Heat in Recuperative Gas Cycle Power Plants: Application to Helium Direct Cycle Nuclear Power Plants

1979 ◽  
Vol 101 (1) ◽  
pp. 130-140 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. F. Jude
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hot Water ◽  
Primary Circuit ◽  
Inlet Temperature ◽  
Total Efficiency ◽  
High Quality ◽  
Quality Process ◽  
Direct Cycle

The advantages of gas turbine power plants in general and closed cycle systems under gas pressure in particular for waste heat recovery are well known. A satisfactory efficiency for electric power generation and good conditions to obtain a significant amount of hot water above 100°C lead to a high fuel utilization. However, as in most of projects, it is not much possible to produce high temperature steam or water without significantly decreasing the electricity production. A new method for an additional generation of high quality process or domestic heat is proposed. The basic feature of this method lies in arranging one or two steam generators or preheaters in parallel with the low pressure side of the recuperator. The high total efficiency and the noteworthy flexibility of this system are emphasized. This arrangement is suitable for any kind of heat source, but the applications presented in this paper are related to helium direct cycle nuclear power plants the main features of which are a single 600 MW(e) turbomachine, a turbine inlet temperature of 775°C, no or one intermediate cooling and a primary circuit fully integrated in a pre-stressed concrete reactor vessel.

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.

Enhanced Verification for RELAP5-3D Parameter and Sensitivity Studies

2016 ◽  
Author(s):  
George L. Mesina ◽  
Nolan Anderson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
User Interaction ◽  
Computer Code ◽  
Unintended Consequences ◽  
Physical Models ◽  
Case Processing ◽  
Time Step ◽  
Sensitivity Studies ◽  
Code Changes

The RELAP5-3D1 program solves a complex system of governing, closure and special process equations to model the underlying physics of nuclear power plants. For SQA (software quality assurance), the code must be verified and validated (V&V) to ensure proper performance before release to users. The physical models are validated against data from experiments and plants and verified against specifications for the computer code. In addition to physics, programs such as RELAP5-3D perform numerous other functions and processes that should also be checked to guarantee correct results. Functions include input, output, data management, and user interaction, while processes include restart, time-step backup, code coupling, and multi-case processing. Previous articles have covered the verification of the physical models, restart, and backup through extremely accurate and automated sequential verification applied on a comprehensive suite of test cases to ensure that code changes produced no unintended consequences. New developments have enabled the verification of multi-case and multi-deck processing. These features are frequently used in parameter and code sensitivity studies and therefore must be verified as working correctly. Both theory and application are presented.

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