scholarly journals Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

2018 ◽  
Vol 2018 ◽  
pp. 1-17
Author(s):  
Frederic Salaun ◽  
David R. Novog
Keyword(s):  
Fuel Cycle ◽  
Control Rod ◽  
Neutron Kinetics ◽  
3 Dimensional ◽  
Thermodynamic Critical Point ◽  
Sensitivity Studies ◽  
Gen Iv ◽  
Safety Systems ◽  
Supercritical Water Cooled Reactor ◽  
Control Rods

The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

scholarly journals Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

2017 ◽  
Vol 2017 ◽  
pp. 1-8 ◽  
Author(s):  
Tagor Malem Sembiring ◽  
Surian Pinem ◽  
Peng Hong Liem
Keyword(s):  
Nuclear Energy ◽  
Neutron Diffusion ◽  
Energy Agency ◽  
Control Rod ◽  
3 Dimensional ◽  
Nodal Method ◽  
Calculation Results ◽  
Control Rods ◽  
Neutron Diffusion Equation ◽  
Good Agreement

The in-house coupled neutronic and thermal-hydraulic (N/T-H) code of BATAN (National Nuclear Energy Agency of Indonesia), NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers), respectively.

scholarly journals GCFR Coupled Neutronic and Thermal-Fluid-Dynamics Analyses for a Core Containing Minor Actinides

2009 ◽  
Vol 2009 ◽  
pp. 1-8 ◽  
Author(s):  
Diego Castelliti ◽  
Eleonora Bomboni ◽  
Nicola Cerullo ◽  
Guglielmo Lomonaco ◽  
Carlo Parisi
Keyword(s):  
Fluid Dynamics ◽  
Energy Production ◽  
Nuclear Power ◽  
Climate Changes ◽  
Fuel Cycle ◽  
Production Of Hydrogen ◽  
Gen Iv ◽  
Safety Systems ◽  
Dynamics Analyses ◽  
Thermal Hydraulic Analysis

Problems about future energy availability, climate changes, and air quality seem to play an important role in energy production. While current reactor generations provide a guaranteed and economical energy production, new nuclear power plant generation would increase the ways and purposes in which nuclear energy can be used. To explore these new technological applications, several governments, industries, and research communities decided to contribute to the next reactor generation, called “Generation IV.” Among the six Gen-IV reactor designs, the Gas Cooled Fast Reactor (GCFR) uses a direct-cycle helium turbine for electricity generation and for a -free thermochemical production of hydrogen. Additionally, the use of a fast spectrum allows actinides transmutation, minimizing the production of long-lived radioactive waste in an integrated fuel cycle. This paper presents an analysis of GCFR fuel cycle optimization and of a thermal-hydraulic of a GCFR-prototype under steady-state and transient conditions. The fuel cycle optimization was performed to assess the capability of the GCFR to transmute MAs, while the thermal-hydraulic analysis was performed to investigate the reactor and the safety systems behavior during a LOFA. Preliminary results show that limited quantities of MA are not affecting significantly the thermal-fluid-dynamics behavior of a GCFR core.

Fuel Cycle Extension by Part-Length Control Rod Removal at Rancho Seco

1981 ◽  
Vol 55 (3) ◽  
pp. 583-586
Author(s):  
T. N. Ake ◽  
R. G. McAndrew ◽  
D. D. Whitney
Keyword(s):  
Fuel Cycle ◽  
Control Rod

Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)

2013 ◽  
Author(s):  
Kazuhiro Kamei ◽  
Kazuyoshi Kataoka ◽  
Kazuto Imasaki ◽  
Noboru Saito
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Plant Availability ◽  
Control Rod ◽  
Water Reactor ◽  
Spray System ◽  
Maintenance And Inspection ◽  
Reactor Pool ◽  
Safety Systems

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide. By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.

Predicting Thermal Mixing and Fatigue Inside Control Rod Guide Tubes

2013 ◽  
Author(s):  
Eric Lillberg
Keyword(s):  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Low Frequency ◽  
Experimental Measurements ◽  
Guide Tube ◽  
Control Rod ◽  
Thermal Mixing ◽  
Purge Flow ◽  
Control Rods ◽  
Good Agreement

The cracked control rods shafts found in two Swedish NPPs were subjected to thermal fatigue due to mixing of cold purge flow with hot bypass water in the upper part of the top tube on which the control rod guide tubes rests. The interaction between the jets formed at the bypass water inlets is the main source of oscillation resulting in low frequency downward motion of hot bypass water into the cold purge flow. This ultimately causes thermal fatigue in the control rod shaft in the region below the four lower bypass water inlets. The transient analyses shown in this report were done to further investigate this oscillating phenomenon and compare to experimental measurements of water temperatures inside the control rod guide tube. The simulated results show good agreement with experimental data regarding all important variables for the estimation of thermal fatigue such as peak-to-peak temperature range, frequency of oscillation and duration of the temperature peaks. The results presented in this report show that CFD using LES methodology and the open source toolbox OpenFOAM is a viable tool for predicting complex turbulent mixing flows and thermal loads.

A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations

2021 ◽  
Author(s):  
Jaakko Leppänen ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Fuel Cycle ◽  
Natural Circulation ◽  
Reactor Core ◽  
District Heating ◽  
Research Centre ◽  
Computational Framework ◽  
Conceptual Design Phase ◽  
Burnable Absorber ◽  
Control Rods ◽  
Type Fuel

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

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.

Estimation of control rod worth, xenon effect on reactivity and power defect of BAEC TRIGA Mark-II research reactor

2018 ◽  
Vol 33 (39) ◽  
pp. 1850233
Author(s):  
Md. Mehedi Hassan ◽  
K. M. Jalal Uddin Rumi ◽  
Md. Nazrul Islam Khan ◽  
Rajib Goswami
Keyword(s):  
Theoretical Analysis ◽  
Low Power ◽  
Temperature Effects ◽  
Research Reactor ◽  
Power Level ◽  
Fuel Temperature ◽  
Control Rod ◽  
Full Power ◽  
Triga Mark Ii ◽  
Control Rods

In this work, control rod worth, xenon (Xe) effect on reactivity and power defect have been measured by doing experiments in the BAEC TRIGA Mark-II research reactor (BTRR) and through established theoretical analysis. Firstly, to study the xenon-135 effect on reactivity, reactor is critical at 2.4 MW for several hours. Next, experiments have been performed at very low power (50 W) to avoid temperature effects. Moreover, for the power defect experiment, different increasing power level has been tested by withdrawing the control rods. Finally, it is concluded that the total control rods worth of the BAEC TRIGA Mark-II research reactor, as determined through this study, is enough to run the reactor at full power (3 MW) considering the xenon-135 and fuel temperature effects.

The Effect of MA Transmutation in the PWR on Fuel Cycle

2017 ◽  
Author(s):  
Haoyang Yu ◽  
Bin Liu ◽  
Wenxin Zhang ◽  
Jin Cai
Keyword(s):  
Fuel Cycle ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Mcnp Program ◽  
And Control ◽  
Control Rods ◽  
Effective Multiplication

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

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