Simulation of Radiation Monitoring Instruments Data for Emergency Response Exercises

2017 ◽  
Author(s):  
Liu Wang ◽  
Liao Feiye ◽  
Jiang Pingting ◽  
He Dongyu ◽  
Luo Yong ◽  
...  
Keyword(s):  
Emergency Response ◽  
Radiation Monitoring ◽  
Severe Accident ◽  
Source Terms ◽  
Simulation Code ◽  
Response Level ◽  
Calculation Results ◽  
Fukushima Daiichi ◽  
Response Exercise

Radiation monitoring instruments (KRT) is important to decide emergency response level (EAL) in accident situation. Emergency response drills is more and more significant after Fukushima Daiichi severe accident. This article develops a code to simulate radiation monitoring instruments data for emergency response exercises. A part of calculated input of the code comes from MAAP calculated results included source terms and thermo hydraulic data. KRT simulation code runs with MAAP calculation results and output KRT simulated value at the same time. This article gives a way to prove the result of KRT simulation code analyzed is correct and matched through simulating an emergency response exercise scenario.

Causes and Radiological Consequences of the Chernobyl and Fukushima Nuclear Accidents

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
L. Sihver ◽  
N. Yasuda
Keyword(s):  
Local Time ◽  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Nuclear Accidents ◽  
Emergency Response System ◽  
Response System ◽  
Accident Management ◽  
Fukushima Daiichi

In this paper, the causes and the radiological consequences of the explosion of the Chernobyl reactor occurred at 1:23 a.m. (local time) on Apr. 26, 1986, and of the Fukushima Daiichi nuclear disaster following the huge Tsunami caused by the Great East Japan earthquake at 2.46 p.m. (local time) on Mar. 11, 2011 are discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating nuclear power plants (NPPs). In addition to that, stress tests should, on a regular basis, be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, decontamination, and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well-prepared and well-established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will also be presented together with the training and education program, which have been established to ensure that professional rescue workers, including medical staff, fire fighters, and police, as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.

Countermeasures to Improve LWR Safety Level in Design and Operation by Accident Analyses at Fukushima Daiichi Nuclear Power Plants

2013 ◽  
Author(s):  
Kazumasa Shimizu ◽  
Yuhei Hamada ◽  
Hiroto Sakashita ◽  
Michitsugu Mori
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Critical Role ◽  
Severe Accident ◽  
Simulation Code ◽  
Safety Level ◽  
Plant Operations ◽  
Fukushima Daiichi ◽  
Water Reactors

The 2011 off the Pacific coast of Tohoku Earthquake occurred on March 11. The earthquake attacked the Fukushima Daiichi nuclear power station with six boiling water reactors (BWRs), three out of which, units 1 through 3 in rated operation except for three reactors of units 4 through 6 in scheduled periodic inspection, automatically shut down in response to the intense seismic motion. Emergency diesel generators started to pump water to cool reactors, and an hour later, the back-up generators lost their all functions by the station blackout resulting from tsunami flooding. In this situation at the unit 1, the isolation condenser system (IC) should have made a critical role to keep the reactor pressure and water level to be safety by removing the decay heat by natural circulation. In fact at the unit 1 during the accident, IC valves were closed by fail-safe and could not have shown the ability of the designed function. An accident report gave general descriptions of the causes and results of accidents, but not the quantitative data indicative of details; therefore, it seems difficult to identify the specific problems in plant operations. Even in this case, if an appropriate analysis code is available for reproducing events based on the reports, it will be possible to determine individual data quantitatively and identify problems in plant operations. In our work, we used the nuclear reactor thermal-hydraulic code RETRAN-3D/MOD4, which has been approved and licensed by U.S. Nuclear Regulatory Commission, to model light water reactors (LWRs) and reproduce the circumstances of the 2011 Fukushima Daiichi nuclear accident as the simulation code. Here, we subjected transition analyses of the process on the core-meltdown accident, and put forward the system to prevent the accident, where the accident analysis report was employed to simulate conditions of the accident. It could enable us to suggest adequate operation procedures suitable for LWR to avoid the severe accident, and to propose countermeasures to improve LWR safety level in design and operation.

Analysis of Accident Progression of Fukushima Daiichi NPPs With SAMPSON Code

2013 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Severe Accident ◽  
Injection System ◽  
Part Load ◽  
Calculation Results ◽  
Fukushima Daiichi ◽  
Measured Pressure

The Fukushima Daiichi Nuclear Power Plant units 1, 2, and 3 had serious damages due to the huge earthquake and tsunami which occurred on March 11th 2011. Pressure transients in the reactor pressure vessels (RPVs) of the units 1, 2, and 3 were analyzed with the severe accident analysis code, SAMPSON for a few days from the scram until occurrence of depressurization. Since preliminary analysis results with the original SAMPSON showed difference from the measured data, the following phenomena were newly considered in the current analyses. For unit 1: Damage of a source range monitor, which is one of in-core monitors. For unit 2: Part load operation of the reactor core isolation cooling system. For unit 3: Part load operation of the high pressure coolant injection system. The calculation results showed fairly good agreements with the measured pressure data and showed RPV bottom damage for all the units resulting in falling of debris in the core region into the pedestal of the drywell.

Development of a Last Stage Blade Row Coupled by Damping Elements: Numerical Assessment of its Vibrational Behavior and its Experimental Validation During Spin Pit Measurements

2017 ◽  
Author(s):  
Christian Siewert ◽  
Frank Sieverding ◽  
William J. McDonald ◽  
Manish Kumar ◽  
James R. McCracken
Keyword(s):  
Structural Integrity ◽  
Mechanical Design ◽  
Vibration Response ◽  
Dynamic Loads ◽  
Steam Turbines ◽  
Response Level ◽  
Vibrational Behavior ◽  
Blade Row ◽  
Calculation Results ◽  
Last Stage

Last stage blade rows of modern low pressure steam turbines are subjected to high static and dynamic loads. The static loads are primarily caused by the centrifugal forces due to the steam turbine’s rotational speed. Dynamic loads can be caused by instationary steam forces, for example. A primary goal in the design of modern and robust blade rows is to prevent High Cycle Fatigue caused by dynamic loads due to synchronous or non-synchronous excitation mechanisms. Therefore, it is important for the mechanical design process to predict the blade row’s vibration response. The vibration response level of a blade row can be limited by means of a damping element coupling concept. Damping elements are loosely assembled into pockets attached to the airfoils. The improvement in the blade row’s structural integrity is the key aspect in the use of a damping element blade coupling concept. In this paper, the vibrational behavior of a last stage blade row with damping elements is analyzed numerically. The calculation results are compared to results obtained from spin pit measurements for this last stage blade row coupled by damping elements.

The Development of Candling Module Code in Module In-vessel Degraded Analysis Code MIDAC and the Relevant Calculation for CPR1000 During Large-Break LOCA

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Jun Wang ◽  
Yuqiao Fan ◽  
Yapei Zhang ◽  
Xinghe Ni ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Mass Distribution ◽  
Pressure Vessel ◽  
Hydrogen Generation ◽  
Peak Temperature ◽  
Severe Accident ◽  
Molten Material ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
Relevant Calculation ◽  
Reactor Pressure

The occurrence of Fukushima has increased the focus on the development of severe accident codes and their applications. As a part of Chinese “National Major Projects,” a module in-vessel degraded analysis code (MIDAC) is currently being developed at Xi’an Jiaotong University. The developing situation of a candling module and relevant calculation for CPR1000 for large break loss of coolant analysis (LOCA) are presented in this paper. The candling module focuses on the melting, moving, and relocation of the melting core materials and necessary thermal hydraulic information. MIDAC’s LOCA accident calculation results of Chinese pressure reactor 1000 (CPR1000) cover the melting mass distribution, peak temperature, and hydrogen generation. The results have been compared with MAAP. Through comparison, the candling module of MIDAC proved to be able to predict the moving trend of the molten material mass relocation in the reactor pressure vessel.

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

Severe accident source terms for a sodium-cooled fast reactor

2014 ◽  
Vol 64 ◽  
pp. 220-229 ◽  
Author(s):  
Acacia Brunett ◽  
Richard Denning ◽  
Marissa Umbel ◽  
Whitney Wutzler
Keyword(s):  
Fast Reactor ◽  
Severe Accident ◽  
Source Terms
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