Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Alexandre de Souza Soares ◽  
Antonio C. M. Alvim

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

Author(s):  
Bo Shi ◽  
Zhao-Fei Tian

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.


Author(s):  
Isaac Malgas

Virtually all operating nuclear plants have a well-developed suite of accident procedures which can be used in the event of an accident while the unit is at power. The procedures guide the operators through optimal recovery strategies to place the plant in a safe stable shutdown state. These procedures cater for design basis accidents as well as certain accidents which are not considered design basis accidents, but may be included in the plant’s licensing basis. Generally procedures used for accidents during power operation can also be used for initiating accidents which occur while the unit is shutdown while the core is cooled with steam generators. When the unit is in lower shutdown states and the residual heat removal system is in service, however, the procedure scope becomes somewhat constrained. With the reactor coolant system in an open or closed configuration, the state/configuration specific procedure guidance provided by Emergency Operating Procedures for power operation is often deficient. There is ample literature on developing emergency operating procedures for plant states where the energy (as a function of pressure and temperature) in reactor coolant system is high. However, much fewer sources of literature exist for shutdown plant states when the energy in the reactor coolant system is lower. This paper attempts to bridge this gap by proposing a framework for the development of shutdown emergency operating procedures using the plant specific Probabilistic Safety Assessment (PSA) model. The paper provides an initial framework for the development of Shutdown Emergency Operating Procedures (SD-EOPs) using the PSA. It describes the process of identifying event sequences for which restoration strategies are required from the PSA and briefly explains the analysis that should be performed to demonstrate that restoration strategies are successful in mitigating event consequences.


Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.


Author(s):  
Nobuyuki Ishikawa ◽  
Yoshitaka Chikazawa ◽  
Daisuke Sato ◽  
Risako Ikari

In this paper, the emergency power supply system for next generation sodium fast reactor (SFR) is investigated to target the latest system design with such as optimized equipment for measure of loss of heat removal system (LOHRS) events, taking into account for strengthening diversity and multiplicity. In addition to diversity and multiplicity, optimization in terms of amount of equipment is also taken into account. The emergency power supply system is consist of emergency generators for design based accident (DBA) and emergency generators to cope with the design extension condition (DEC) in order to power the additional cooling equipment of fuel handling system and reactor core cooling system as a measure for LOHRS event. Base on the load capacity and system operation, the configuration of emergency power supply system was examined aiming optimization of amount of equipment in accordance with some shared use taking into account the reliability. As a result, the emergency power supply system with 4 fixed-type gas turbine generators, 4 portable-type diesel generators and batteries was established.


Author(s):  
Edward L. Carlin ◽  
Peter A. Hilton ◽  
Yixing Sung

The Reactor Coolant System (RCS) of the AP1000 plant consists of two circulating loops. Each loop contains two canned motor Reactor Coolant (RC) pumps that have a rotating inertia to provide RCS flow coastdown if power to the pumps is lost. Westinghouse analysis of the complete loss of flow (CLOF) accident in support of the AP1000 design certification was based on the USNRC-approved traditional methodology applied to operating plants. The RCS response during the transient was predicted using the LOFTRAN code based on a reactivity insertion curve highly skewed to the bottom of the reactor core, but the calculation of Departure from Nucleate Boiling Ratio (DNBR) was performed assuming a top-skewed axial power profile. A more realistic margin assessment can be made by using an improved method similar to Westinghouse RAVE methodology recently approved by the USNRC. The improved method uses the three-dimensional kinetic nodal code SPNOVA coupled with the reactor core thermal-hydraulic code VIPRE-W for predicting the reactor core response during the CLOF transient. The improved method significantly improves margin predictions by generating core power distributions consistent with the trip reactivity changes for the DNBR calculation. The margin assessment showed that the improved method resulted in a 19% DNBR increase as compared to the traditional method for the AP1000 CLOF transient.


2008 ◽  
Vol 73 (10) ◽  
pp. 1340-1356 ◽  
Author(s):  
Katarína Mečiarová ◽  
Laurent Cantrel ◽  
Ivan Černušák

This paper focuses on the reactivity of iodine which is the most critical radioactive contaminant with potential short-term radiological consequences to the environment. The radiological risk assessments of 131I volatile fission products rely on studies of the vapour-phase chemical reactions proceeding in the reactor coolant system (RCS), whose function is transferring the energy from the reactor core to a secondary pressurised water line via the steam generator. Iodine is a fission product of major importance in any reactor accident because numerous volatile iodine species exist under reactor containment conditions. In this work, the comparison of the thermodynamic data obtained from the experimental measurements and theoretical calculations (approaching "chemical accuracy") is presented. Ab initio quantum chemistry methods, combined with a standard statistical-thermodynamical treatment and followed by inclusion of small energetic corrections (approximating full configuration interaction and spin-orbit effects) are used to calculate the spectroscopic and thermodynamic properties of molecules containing atoms H, O and I. The set of molecules and reactions serves as a benchmark for future studies. The results for this training set are compared with reference values coming from an established thermodynamic database. The computed results are promising enough to go on performing ab initio calculations in order to predict thermo-kinetic parameters of other reactions involving iodine-containing species.


Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.


Author(s):  
Shoji Takada ◽  
Shunki Yanagi ◽  
Kazuhiko Iigaki ◽  
Masanori Shinohara ◽  
Daisuke Tochio ◽  
...  

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.


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