Reduction of Dynamic Response of Buildings, Equipment and Pipe Networks Using SERB-SITON Method to Encrease NPP Safety Margins

Author(s):  
Viorel Serban ◽  
Adrian Panait ◽  
Marian Androne ◽  
George Alexandru Ciocan ◽  
Madalina Zamfir

Buildings, equipment and pipe networks, herein below called “structures”, within nuclear and classic objectives are affected by the dynamic actions of earthquake, shocks and vibrations type, herein below called“ excitations”. The repeated action of excitations on structures most often lead to important built-up of kinetic and potential energy in oscillating systems made-up of such structures. Such a built-up is leading to tens of times increase of the amplitude of the dynamic response in accelerations and displacements of structures as to the excitation amplitude. In its turn, such an increase is leading to the exceeding of the efforts and distortions of the structures material, accompanied by the occurrence of damages or destroy. Because in most cases the excitation cannot be reduced or eliminated, the only technical solution available for the engineers is to reduce the dynamic response of the structures on the existing excitations. The innovative SERB-SITON solution developed and applied within the Subsidiary of Technology and Engineering for Nuclear Projects (SITON) is based on the use of SERB-SITON isolation behavior devices with low friction and controlled stiffness on horizontal plane to cut-off the dynamic action transfer from the excitation to the structure and telescopic devices with controlled elasticity and damping both to reduce the dynamic response of the structures, as well as the dissipation of the energy transferred to the structure or the limitation of structure distortion upon the dynamic actions. The paper presents the performances of SERB-SITON mechanical devices used to protect buildings, equipment and pipe networks against dynamic actions and an adequate method to evaluate the dynamic response of structures (by using SERB-SITON devices), in real time inclusively. For large-size buildings such as nuclear power plants, old buildings, churches, bridges, etc., SERB-SITON isolation devices with friction by sliding or rolling with a very small and adjustable, stiffness installed under the structure on any horizontal direction, are presented. For reinforced concrete or metal framework buildings such as high buildings, industrial halls, towers, etc., SERB-SITON telescopic devices with controlled stiffness and damping large, force inclusively are presented. For equipment and pipe networks, SERB-SITON supports that are capable to overtake large permanent loads with relative displacements on two directions for thermal displacements, and also capable to elastically overtake and damp dynamic actions, are presented.

Author(s):  
Xiaoyao Shen ◽  
Yongcheng Xie

The control rod drive mechanism (CRDM) is an important safety-related component in the nuclear power plant (NPP). When CRDM steps upward or downward, the pressure-containing housing of CRDM is shocked axially by an impact force from the engagement of the magnetic pole and the armature. To ensure the structural integrity of the primary coolant loop and the functionality of CRDM, dynamic response of CRDM under the impact force should be studied. In this manuscript, the commercial finite element software ANSYS is chosen to analyze the nonlinear impact problem. A nonlinear model is setup in ANSYS, including main CRDM parts such as the control rod, poles and armatures, as well as nonlinear gaps. The transient analysis method is adopted to calculate CRDM dynamic response when it steps upward. The impact loads and displacements at typical CRDM locations are successfully obtained, which are essential for design and stress analysis of CRDM.


Author(s):  
Manel Ellouz ◽  
Eva Kasparek ◽  
Holger Völzke

Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and “old” steel sheet box shaped ones. The so called “old” containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM.


Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet ◽  
Georges Bezdikian

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the micro-structural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability. This paper presents several tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on two-third scale elbows and welded mock-ups. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized. These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.


Author(s):  
Luben Sabotinov ◽  
Abhishek Srivastava ◽  
Pierre Probst

In the accident analysis of the Nuclear Power Plants (NPP) nowadays the international licensing practice considers several acceptable options for demonstrating the safety i.e. use of conservative computer codes with conservative assumptions, best estimate codes combined with conservative assumptions and conservative input data and application of best estimate codes with assumptions and realistic input data but associated with uncertainty evaluation of the results. The last option is particularly attractive because it allows for more precise prediction of safety margins with respect to safety criteria and their future use for power up-rating. The best estimate simulation with uncertainty analysis constitutes the framework of the present study which is to apply the last version of the French best estimate computer code CATHARE 2 in order to predict the thermal-hydraulic phenomena in the Indian KudanKulam Nuclear Power Plant (KK NPP) with VVER-1000 reactors during LB LOCA and to evaluate uncertainty along with sensitivity studies using the IRSN methodology. The paper first describes the modeling aspects of LB LOCA with CATHARE and then it presents the basic results. It highlights the use of SUNSET statistical tool developed by IRSN for sampling, management of several runs using CATHARE and further post treatment for uncertainty and sensitivity evaluation. The paper also deals with the difficulties associated with the selection of input uncertainties, code applicability and discusses the challenges in uncertainty evaluation.


Author(s):  
Shakir Y. Haider ◽  
David J. Calhoun

After the 2011 Japan earthquake and tsunami caused significant damage at Fukushima Daiichi, the Nuclear Regulatory Commission required all US nuclear power plants to have a mitigation strategy for beyond design basis events. Industry-developed response plans, called “FLEX” strategies, deployed new, portable equipment such as diesel generators and cooling pumps. As this new equipment needs to be available after a natural disaster, storage in protected locations is required. Many nuclear plants have recently constructed new storage buildings, or FLEX buildings, as part of their post-Fukushima strategy. The equipment door is a critical component of a FLEX building. Large enough to drive a semi-trailer truck through, it is required to protect the equipment in case of an earthquake, flood, tornado and also may need to be capable of opening within a few minutes in order to respond during an emergency. The equipment door presented in this paper serves these purposes very effectively. The composite section of the door is capable of shielding the structure from penetration as well as overall dynamic response from tornado missile impact. The door travels on an overhead rail which, being indoors and above the opening, provides reliable door movement in case of snow or ice during winters or in case of debris from wind or tornado. Latches capable of withstanding tornado missile impact forces also restrain the door in case of wind or seismic forces. The door is opened by means of motorized trollies and is also equipped with a backup opening device by means of an air winch in case of a power loss. The door and the latches that restrain the door from impact are analyzed using ANSYS finite element software. A limit state analysis is performed that identifies the sequence of yield limit states for the components of the door and the door latch as the loading progresses. The analysis continues until the ductility limit state for the latch is reached. Redistribution of stresses within the components of the latch is observed during the analysis. A modal analysis and a direct integration time history analysis is also performed to capture the dynamic response due to impact loading. Overall, this paper presents a highly robust and reliable design for a FLEX building equipment door that is capable of protecting the contents of the building during a natural disaster and remaining operational during the response after an emergency.


Author(s):  
M. Doucet ◽  
P. Faye ◽  
Ch. Faignet ◽  
R. Babut ◽  
M. Landrieu ◽  
...  

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty, • …. The French criticality code package CRISTAL is used to check several configurations reactivity and derived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA® resin as neutron absorber. The cask is designed to handle mainly the EPR™ fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the applied uncertainties will be discussed.


Author(s):  
L. Ike Ezekoye ◽  
William E. Densmore ◽  
William M. Turkowski ◽  
Robert E. Becse

Check valves are the simplest valves in power plants. Their simplicity and passive nature, combined with their relatively low maintenance requirements, often mask their relative importance in piping systems. Compared to power operated valves (POVs), such as motor operated valves or air operated valves, check valves have very few parts. The more parts a valve has, the more likely failures will occur. As such, power operated valves tend to have more stringent requirements that cover periodic verification of operability, in-service testing (IST), and scheduled preventive maintenance to assure functionality. Check valves, on the other hand, do not require nearly the same amount of rigor to assure operability. The passive nature of check valves often leads the user not to expect failures. Consequently, lacking of attention often results to inadvertent failures. One failure that has received significant attention from both the industry and the regulator is check valve body-to-bonnet joint leakage. In nuclear power plants this leakage can contaminate the general area where the valve is located, can lead to a plant shutdown, and pose personnel hazards. In this paper, the technical solutions that can be used to manage check valve body-to-bonnet joint leakage will be presented. The merits of each technical solution and the associated challenges will be discussed. Also, as some of the leakage containment solutions are appurtenances to the valve, the paper will address the interface between the appurtenances and the valve.


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