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NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient.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 The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Surian Pinem ◽  
Tagor Malem Sembiring ◽  
Peng Hong Liem
Keyword(s):  
Coolant Temperature ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
3 Dimensional ◽  
Nodal Method ◽  
Benchmark Calculation ◽  
Power Peak ◽  
Calculation Results ◽  
Neutron Diffusion Equation

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised). Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

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.

Study on Semi-Analytical Nodal Method for Solving SP3 Equation

2017 ◽  
Author(s):  
Pan Qingquan ◽  
Lu Haoliang ◽  
Li Dongsheng ◽  
Wang Kan
Keyword(s):  
Flux Distribution ◽  
High Accuracy ◽  
Neutron Diffusion ◽  
Expansion Process ◽  
Reactor Physics ◽  
Key Technology ◽  
Nodal Method ◽  
Order Algebraic ◽  
The Difference ◽  
Neutron Diffusion Equation

Solving the SP3 equation is the key technology of the Next Generation Reactor Physics Calculation, and has been widely concerned. The semi-analytical nodal method (SANM) based on transverse-integrated neutron diffusion equation has the advantages of high accuracy and convenience for multi-group calculation. The 0th-order flux and the 2nd-order flux being Expanded with the existing 4th-order SANM polynomials and being solved respectively, the 4th-order algebraic accuracy flux distribution is also obtained, however, this solving process is not the semi-analytical nodal method since the polynomial expansion process does not take the special modality of SP3 equation and it’s analytical solution into consideration. There are two modality SP3 equation, so there are two SANM expansion forms. A code is developed to solve the SP3 equation under the two different forms. After the calculation of the same benchmark, the difference between the two forms of SP3 equation is found. According to the results, and in view of the special modality of the SP3 equation, advices for a strict semi-analytical method for solving SP3 equation are discussed.

scholarly journals The effects of fuel type on control rod reactivity of pebble-bed reactor

Nukleonika ◽  
2019 ◽  
Vol 64 (4) ◽  
pp. 131-138
Author(s):  
◽  
Topan Setiadipura ◽  
Jim C. Kuijper ◽  
Keyword(s):  
Reactor Core ◽  
Fuel Type ◽  
Reactor Model ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Control Rod ◽  
Reactor Geometry ◽  
Calculation Results ◽  
Control Rods ◽  
Selection Of

Abstract As a crucial core physics parameter, the control rod reactivity has to be predicted for the control and safety of the reactor. This paper studies the control rod reactivity calculation of the pebble-bed reactor with three scenarios of UO2, (Th,U)O2, and PuO2 fuel type without any modifications in the configuration of the reactor core. The reactor geometry of HTR-10 was selected for the reactor model. The entire calculation of control rod reactivity was done using the MCNP6 code with ENDF/B-VII library. The calculation results show that the total reactivity worth of control rods in UO2-, (U,Th)O2-, and PuO2-fueled cores is 15.87, 15.25, and 14.33%Δk/k, respectively. These results prove that the effectiveness of total control rod in thorium and uranium cores is almost similar to but higher than that in plutonium cores. The highest reactivity worth of individual control rod in uranium, thorium and plutonium cores is 1.64, 1.44, and 1.53%Δk/k corresponding to CR8, CR1, and CR5, respectively. The other results demonstrate that the reactor can be safely shutdown with the control rods combination of CR3+CR5+CR8+CR10, CR2+CR3+CR7+CR8, and CR1+CR3+CR6+CR8 in UO2-, (U,Th)O2-, and PuO2-fueled cores, respectively. It can be concluded that, even though the calculation results are not so much different, however, the selection of control rods should be considered in the pebble-bed core design with different scenarios of fuel type.

scholarly journals Deconvolution methods used for the development of a neutron spectrometer

2021 ◽  
Vol 253 ◽  
pp. 09003
Author(s):  
Claude-Alexandre Simonetti ◽  
Marc Labalme ◽  
Jean-Lionel Trolet ◽  
Patrick Mary
Keyword(s):  
Neutron Energy ◽  
Nuclear Energy ◽  
Incident Neutron ◽  
Neutron Spectrometer ◽  
Neutron Detectors ◽  
Energy Agency ◽  
Bonner Sphere ◽  
Computer Codes ◽  
Different Depths ◽  
Good Agreement

To fulfill the requirements of the industry, the feasibility of a transportable neutron spectrometer is under study by our collaboration. Preliminary studies have led to a solution based on a unique-multi-detectors-Bonner sphere, which has the advantage to simultaneously perform many neutrons-measurements through thermal neutron detectors placed at different depths inside a polyethylene sphere. The incident neutron energy is then reconstructed using unfolding methods. The optimization of the layout of the detectors in the sphere and of the diameter of the sphere was performed thanks to GEANT4 simulations, coupled to unfolding methods such as least-squares, maximum of entropy or maximum of likelihood. Different unfolding methods have been tested. For the time being, the best unfolding results were obtained by the computer codes M.A.X.E.D. and G.R.A.V.E.L., formalized by the Nuclear Energy Agency (N.E.A.). A “personal” method based on the maximum of entropy coupled to the maximum of likelihood gives good results as well, but a few convergence parameters have still to be optimized. In the present paper, a solution of a multi-detectors-Bonner sphere is presented as well as results of unfolded neutron energy spectra. The results obtained show a good agreement with unmoderated and moderated Am/Be and 252Cf spectra.

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.

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