scholarly journals Multiphysics Modeling of Thorium-Based Fuel Performance With Cr-Coated SiC/SiC Composite Under Normal and Accident Conditions

2021 ◽  
Vol 9 ◽  
Author(s):  
Shengyu Liu ◽  
Rong Liu ◽  
Chengjie Qiu ◽  
Wenzhong Zhou
Keyword(s):  
Failure Time ◽  
Mechanical Performance ◽  
Hoop Strain ◽  
Fuel Performance ◽  
Multiphysics Modeling ◽  
Normal Operating ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Work First ◽  
Centerline Temperature

Using the finite element multiphysics modeling method, the performance of the thorium-based fuel with Cr-coated SiC/SiC composite cladding under both normal operating and accident conditions was investigated in this work. First, the material properties of SiC/SiC composite and chromium were reviewed. Then, the implemented model was simulated, and the results were compared with those of the FRAPTRAN code to verify the correctness of the model used in this work. Finally, the fuel performance of the Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel, and UO2 fuel combined with the Cr-coated SiC/SiC composite cladding and Zircaloy cladding, respectively, was investigated and compared under both normal operating and accident conditions. Compared with the UO2 fuel, the Th0.923U0.077O2 and Th0.923Pu0.077O2 fuels were found to increase the fuel centerline temperature under both normal operating and reactivity-initiated accident (RIA) conditions, but decrease the fuel centerline temperature under loss-of-coolant accident (LOCA) condition. Moreover, compared to the UO2 fuel with the Zircaloy cladding, thorium-based fuels with Cr-coated SiC/SiC composite cladding were found to show better mechanical performance such as delaying the failure time by about 3 s of the Cr-coated SiC/SiC composite cladding under LOCA condition, and reducing the plenum pressure by about 0.4 MPa at the peak value in the fuel rod and the hoop strain of the cladding by about 16% under RIA condition.

Extending the application range of a fuel performance code from normal operating to design basis accident conditions

2008 ◽  
Vol 383 (1-2) ◽  
pp. 137-143 ◽  
Author(s):  
P. Van Uffelen ◽  
C. Győri ◽  
A. Schubert ◽  
J. van de Laar ◽  
Z. Hózer ◽  
...  
Keyword(s):  
Fuel Performance ◽  
Application Range ◽  
Normal Operating ◽  
Design Basis ◽  
Design Basis Accident ◽  
Accident Conditions

Best-Estimate Plus Uncertainty Analysis of CANDU Fuel Reliability Using Manufacturing and Simulated Core Data

2020 ◽  
Vol 6 (4) ◽  
Author(s):  
Jason J. Song ◽  
Paul K. Chan ◽  
Hugues W. Bonin ◽  
Mahesh Pandey
Keyword(s):  
Internal Pressure ◽  
Probability Distributions ◽  
Real Data ◽  
Simulation Method ◽  
Failure Criteria ◽  
Hoop Strain ◽  
Fuel Performance ◽  
Best Estimate ◽  
Average Improvement ◽  
Centerline Temperature

Abstract A novel method of assessing the reliability of 37-element Canada deuterium uranium (reactor) (CANDU) fuel bundle was explored. The method implements a “best-estimate plus uncertainty” (BEPU) approach where a probabilistic treatment of manufacturing and operating inputs is used to predict fuel performance. The fuel performance was predicted using the Canadian industry standard codes for fuel performance, ELESTRESS and ELOCA, which, respectively, model fuel behaviors during normal and transient conditions. The outputs of the codes were compared against failure criteria from industry norms to determine the probability of failure. A Monte Carlo simulation method was applied to analyze this problem. Probability distributions of manufacturing input variables were estimated from real data, which were then randomly sampled. The inputs for fuel burnup and power were simulated using core-following data generated using a three-dimensional diffusion code, the Reactor Fuelling Simulation Program (RFSP), which were also then randomly sampled. The results of the simulations predict significant improvements in margins to limits for all performance parameters. An average improvement of 500 °C in centerline temperature, 10 °C in sheath temperature, 12 MPa in element internal pressure, and 0.8% in pellet end sheath hoop strain was predicted for the highest-powered region of the core, during normal operations, in comparison with the limit-of-envelope (LOE) benchmark. An 80% reactor overhead break (ROH) transient simulation was also simulated, and an average improvement of 500 °C in centerline temperature, 150 °C in sheath temperature, 6.5 MPa in internal pressure, and 2% in sheath hoop strain was predicted.

Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident

2020 ◽  
Author(s):  
Chenjie Qiu ◽  
Rong Liu ◽  
Wenzhong Zhou
Keyword(s):  
Steady State ◽  
Safety Margin ◽  
Operating Conditions ◽  
Fuel Performance ◽  
Transient Conditions ◽  
Normal Operating ◽  
Accident Conditions ◽  
Behavior Models ◽  
Line Temperature ◽  
Thermal Physical Properties

Abstract The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed. And then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (RIA and LOCA) are analyzed and compared. The Th0.923U0.077O2 fuel is found to have lower fuel center-line temperature and the thorium-based fuels are observed to have a delayed pellet-cladding mechanical interaction (PCMI) under steady state. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than conventional UO2 fuel under both normal operating conditions and accident conditions.

Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT

2018 ◽  
Author(s):  
Jian Li ◽  
Ding She ◽  
Lei Shi ◽  
Jing Zhao
Keyword(s):  
Performance Analysis ◽  
Mechanical Performance ◽  
Performance Model ◽  
Fuel Performance ◽  
Triso Fuel ◽  
Coated Particle ◽  
First Results ◽  
Fuel Particles ◽  
Accident Conditions ◽  
Radioactivity Retention

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Modification of Filtered Air Intake Flowrate to Improve Control Room Habitability

2018 ◽  
Author(s):  
Xinjian Liu ◽  
Weipeng Shu ◽  
Mengxi Wang
Keyword(s):  
Extended Period ◽  
Leakage Rate ◽  
Design Parameters ◽  
Source Terms ◽  
Control Room ◽  
Limit Operator ◽  
Radioactive Aerosols ◽  
Loss Of Coolant Accident ◽  
Operator Dose ◽  
Accident Conditions

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc. Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition. This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation. This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.

Sign in / Sign up

Export Citation Format

Share Document