Evaluation of Alternate Water Injection Cooling for Accident Conditions in Spent Fuel Pool Using MAAP Code

2019 ◽  
Vol 2019 (0) ◽  
pp. S08106
Author(s):  
Satoshi NISHIMURA ◽  
Masaaki SATAKE ◽  
Yoshihisa NISHI ◽  
Yoshiyuki KAJI ◽  
Yoshiyuki NEMOTO
Keyword(s):  
Water Injection ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Accident Conditions

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

The Importance of Reliable Indications for Accident Management

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
Bill T. Williamson
Keyword(s):  
Severe Accident ◽  
Spent Fuel ◽  
Critical Importance ◽  
Dc Power ◽  
The Status ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
And Control ◽  
Parameter Values ◽  
Support Guidelines

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. There is evidence that the failure of key instrumentation to provide reliable information to the control room licensed operators contributed to the severity of the accident at both TMI and Fuskushima Daiichi. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data and yet have to make urgent decisions. While progress in these areas has been made since TMI-2, the accident at Fukushima suggests there may still be some potential for further improvement in critical plant instrumentation. As a result, several approaches are being employed to provide better information to emergency response personnel during a severe accident. The first approach being taken by the PWROG and BWROG is the identification of methods to obtain information related to key plant parameters when there is a loss of dc power for instrumentation and control. The FLEX guidance in NEI 12-06 requires that reliable instrumentation be available to ensure core, containment and spent fuel pool cooling is maintained for the beyond design basis events for which FLEX was intended. For the most part, this instrumentation that is important for FLEX is the same instrumentation that is used for diagnosis of severe accident conditions and challenges to fission product barriers. Generic FLEX Support Guidelines have been developed to provide a uniform basis for plants to meet the NEI 12-06 requirements that includes methods to obtain key parameter values in the event of a loss of all dc instrument power. The PWROG and the BWROG have also taken a complimentary approach to provide Technical Support Guidance (TSG) for instrumentation during a severe accident. This approach identifies the primary instrumentation as well as alternate instrumentation and other tools to validate the indications from the primary instrumentation. The validation consists of: a) comparing the primary instrument indications to the alternate instrumentation, b) comparing instrument indications to related instrumentation, c) comparing instrument indications and trends to expected trends based on the accident progression and actions already implemented, and d) comparing instrument indications to information in calculational aids.

Behavior of the Spent Fuel Pool and Dose Rate Calculations for the CPR1000 Reactor Under Severe Accident Conditions Using the Integral Code MELCOR

2013 ◽  
Author(s):  
Klaus Mueller ◽  
Moses Yeung ◽  
Justin Byard ◽  
Zhen Xun Peng ◽  
Jun Tao ◽  
...  
Keyword(s):  
Fission Product ◽  
Dose Rate ◽  
Flow Rate ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Pure Hydrogen ◽  
Spent Fuel ◽  
Product Release ◽  
Spent Fuel Pool ◽  
Accident Conditions

The behavior of the spent fuel pool and the fission product release and transport for the CPR1000 reactor under severe accident conditions was analyzed using the integral severe accident code MELCOR. In the investigated accident scenario a total failure of the pump of the spent fuel cooling system was assumed. Furthermore, it is assumed that accident management fails to bring water into the spent fuel pool using mobile pumps or due to the non-recovery of the cooling pump. The grace time available for measures in order to avoid significant fission product release to the environment is determined. The calculated hydrogen mass flow rate due to clad oxidation and the steam flow rate from the spent fuel pool to the compartment above the spent fuel pool serve as boundary conditions for the three dimensional fluid dynamics code GASFLOW to assess possible hydrogen combustion or detonation in the compartment. Using this spent fuel pool MELCOR model the dose submerged in air or water can be determined. The calculated gamma dose rate in a specific compartment can be used for equipment qualification and compartment accessibility assessment. It was found that after four days the fuel assemblies are significantly heated-up and ten hours later the fission products are released as well as a significant amount of hydrogen is produced. A preliminary GASFLOW analysis shows by assuming an air atmosphere in the fuel building, that the risk of a hydrogen combustion or detonation is high. In late state of the accident a convection flow of pure hydrogen is established in spent fuel pool region. It was shown, that the flow conditions strongly influence the fission product transport behavior and consequently the dose rates in the compartment above the spent fuel pool.

Calculation of Evaporation From Fukushima NPP Spent Fuel Pools

2018 ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Safety Systems ◽  
Good Agreement

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and post-accident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The post-accident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25 % lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Spent Fuel Pool Accidents in a Nordic BWR

2013 ◽  
Author(s):  
Niina E. Könönen
Keyword(s):  
Natural Circulation ◽  
Water Injection ◽  
Heat Power ◽  
Spent Fuel ◽  
Fuel Temperature ◽  
Fuel Rods ◽  
Decay Heat ◽  
Fuel Assemblies ◽  
Loss Of Coolant ◽  
Spent Fuel Pool

Loss-of-pool-cooling accidents at the spent fuel pools in the reactor hall of a Nordic BWR have been studied using the MELCOR 1.8.6 code. Studies were made with several different calculation nodalizations. Other investigated variables were the total decay heat power of fuel assemblies in the pool, the initiator of the accident (loss of pool cooling or loss of coolant from the pool), the LOCA leak elevation, the alignment of the re-flooding injection and the use of a lid on top of the pool. From the results it was observed that ensuring natural circulation of air in the fuel is essential in preventing fuel damages. In cases where the air flow is prevented (loss of pool cooling, LOCA break elevation above the bottom of fuel assemblies or a lid on top of the pool) the fuel is damaged in nearly all cases. Only with the pool decay power being sufficiently low (2227 kW) the melting of the fuel was prevented. If water injection to the SFP can be restarted before the fuel temperature reaches the cladding failure criterion (1173 K), the rewetting of fuel assemblies is successful and the fuel temperatures will quickly decrease and the assemblies will remain intact. The re-flooding of already damaged fuel assemblies will also cool the fuel rods and reduce the radioactive releases to the environment. However, it may result in greater hydrogen releases.

Thermal Hydraulics Analysis on a Simplified Model of Spent Fuel Pool

2013 ◽  
Author(s):  
Min Liang ◽  
Daogang Lu ◽  
Huining Xia ◽  
Yuhao Zhang
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Cooling Water ◽  
Measuring Instruments ◽  
Spent Fuel ◽  
Circulating Water ◽  
Experimental Process ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Hydraulic Process

The spent fuel still has massive decay heat and is temporarily stored in the spent fuel pool, after unloaded from the reactor core; it is cooled by the circulating water in the spent fuel pool. The present study generally agreed that the spent fuel pool is relatively safe when the cooling water works, even if under accident conditions. It is generally considered that the thermal-hydraulic process in the spent fuel pool is very slow, and does not endanger the safety of nuclear power plant. But the boiling in the pool has been given more and more attention after the Fukushima nuclear accident; especially after the failure of core safety injection and the cooling water in the spent fuel pool is lost. School of Nuclear Science and Engineering, North China Electric Power University intends to perform overall test bench to simulate the AP1000 spent fuel pool, the experimented data can be used to validate the results of the COSINE software calculations. In this paper, the numerical simulation is performed with CFD software to calculate the temperature, velocity, pressure field of the scaled spent fuel pool experimental bench; it can provide the guideline for the selection of the test section measuring instruments and the arrangement of the measuring points. Also, it’s essential to verify the COSINE software and guide the experimental process.

The holistic risk analysis of spent fuel pool considering the fuel route risk

2021 ◽  
pp. 1748006X2199851
Author(s):  
Zhixin Xu ◽  
Ming Wang ◽  
Binyan Song ◽  
WenYu Hou ◽  
Chao Wang
Keyword(s):  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Point Of View ◽  
Spent Fuel ◽  
Nuclear Disaster ◽  
Evaluation Approach ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Large Water ◽  
Fukushima Nuclear Disaster

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

Influence of Gap Size on Added Mass for Spent Fuel Storage Rack

2018 ◽  
Author(s):  
Daogang Lu ◽  
Yu Liu ◽  
Shu Zheng
Keyword(s):  
Vibration Frequency ◽  
Input Parameter ◽  
Added Mass ◽  
Water Tank ◽  
Spent Fuel ◽  
Free Standing ◽  
Fluid Force ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Spent Fuel Pool

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1]. The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool. In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.

Sign in / Sign up

Export Citation Format

Share Document