scholarly journals BWR/5 Pressure-Suppression Pool Response during an SBO

2013 ◽  
Vol 2013 ◽  
pp. 1-8
Author(s):  
Javier Ortiz-Villafuerte ◽  
Andrés Rodríguez-Hernández ◽  
Enrique Araiza-Martínez ◽  
Luis Fuentes-Márquez ◽  
Jorge Viais-Juárez
Keyword(s):  
Pressure Rise ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Power Station ◽  
The Core ◽  
Initial Event ◽  
Station Blackout ◽  
Suppression Pool ◽  
Failure Temperature ◽  
Design Pressure

RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Deposition Fractions of Fission Product and Heavy Elements on Soil in Fukushima Dai-Ichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Toru Yamamoto
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Soil Samples ◽  
Heavy Elements ◽  
Severe Accident ◽  
Accident Analysis ◽  
Radioactivity Measurement ◽  
Power Station ◽  
The Core

Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10−4 to 10−3 for Sr, 10−4 for Nb, 10−2 to 10−1 for Mo, 10−1 for Ag, 10−1 to 100 for Te, 100 for I, 10−3 for Ba, 10−6 to 10−5 for Pu, 10−6 to 10−5 for Am, and 10−6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis.

LBLOCA Initiated Emergency Condition Analysis for a China Three-Loop PWR

2018 ◽  
Author(s):  
Wang Ning ◽  
Chen Lei ◽  
Zhang Jiangang ◽  
Yang Yapeng ◽  
Xu Xiaoxiao ◽  
...  
Keyword(s):  
Severe Accident ◽  
Feed Water ◽  
Fukushima Accident ◽  
Flow Paths ◽  
The Core ◽  
Emergency Condition ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Control Volumes

Great interest in severe accident has been motivated since Fukushima accident, which indicates that the probability of severe accident exists even though it is extremely small. Emergency condition is important in decision making in case of severe accident in NPP. Although many studies have been conducted for severe accident, there was necessary to investigate emergency condition of severe accidents that could possibly happen and haven’t been sufficiently analyzed. Since station blackout (SBO) happened in Fukushima accident, a number of studies in severe accidents initiated by SBO have been carried out. Off-site power is assumed to be lost during large break loss of coolant accident (LBLOCA), but there is few study to find out emergency condition during LBLOCA if both of off-site and on-site power are lost. A hypothetical severe accident initiated by LBLOCA along with SBO in a China three-loop PWR was simulated in the paper using MELCOR code. Emergency condition was obtained including start of core uncover, start of zirconium-water reaction, failure of fuel cladding and failure of the lower head. Thermal-hydraulic response of the core during the accident was also analyzed in the paper. The model for this study consists of 46 control volumes (27 in primary loop, 17 in secondary loop, 1 in containment and 1 in environment) and 52 flow paths. High pressure safety injection (HPSI) and low pressure safety injection (LPSI) are lost because of loss of on-site and off-site power, and simultaneously main feed water and auxiliary feed water of the steam generators are lost for the same reason. The accumulator can inject water into the core since it is passive and doesn’t need any power. Results of the study will be useful in gaining an insight into detailed severe accident emergency condition that could happen in a China three-loop PWR and may provide basis for severe accident mitigation.

Challenges in Multi-Unit CANDU Reactor Severe Accident Mitigation Strategies

2016 ◽  
Author(s):  
Sunil Nijhawan
Keyword(s):  
Analytical Methods ◽  
Public Safety ◽  
Severe Accident ◽  
Regulatory Regime ◽  
Steam Generators ◽  
Management Options ◽  
Candu Reactors ◽  
Accident Management ◽  
Station Blackout ◽  
Design Pressure

While most of the severe accident related vulnerabilities arising from the inherent 40 odd year old PHWR design are common with single unit CANDU reactors and a number are also shared with LWR designs of that vintage, an evaluation of a station blackout accident at a multi-unit CANDU station reveals significant challenges to accident management options and potentially unacceptable off site radiological consequences. Opportunities for design improvements are abundant but unfortunately mostly ignored with both accident progression and consequence assessments by the utilities presented in a distorted positive light in defiance of engineered realities and public safety. Over-pressure protection systems in all relevant reactor systems (PHTS, Calandria, Shield Tank, and Containment) are inadequate for decay heat, let alone for other anticipated severe accident loads. Early passive heat removal by steam generators after a station blackout can be compromised by primary coolant removal into a large pressurizer located well below the pump bowl. There are no emergency means of high pressure water addition to the steam generators or the heat transport system which not only has an inadequate steam relief capacity for over pressure protection such that an early containment bypass by steam generator tube ruptures is a possibility, but also lacks a method of manual depressurization for early accident mitigation. In absence of a retaining LWR like pressure vessel, the reactor cores would release fission products without attenuation into the box like containments that are at 48% per day leak rate at design pressure very leaky and at less than 1 bar design pressure, structurally weakest of all operating reactor containments. The reactor buildings around each individual reactor unit are inverted cup like traps for combustible gases. A large number of safety significant components like the steam generators, pumps and the reactivity control devices are all outside the containment envelope. The production of combustible Deuterium gas from over ten km of carbon steel piping and over 50 tons of Zircaloy can be extremely high making the installed numbers and types of PARS not only inadequate but as early ignition sources also dangerous. Improvements after Fukushima are perfunctory and the analytical methods in support of severe accident management guidelines are outdated and incomplete. A lax and uninformed regulatory regime blindly supporting an intransigent industry resisting basic design enhancements has further exasperated, like it did in Japan, the severe accident related risk from continued operation of these reactors. These conclusions are based on thirty years of working on severe accident related issues at CANDU reactors, conducting extensive design reviews and developing computer codes and analytical methods for accident progression and consequence assessments. It is hoped that open discussions by professional engineers would foster change in name of public safety. It is also feared that nothing will change unless an accident occurs.

Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC

2009 ◽  
Author(s):  
Polina Tusheva ◽  
Nils Reinke ◽  
Eberhard Altstadt ◽  
Frank Schaefer ◽  
Frank-Peter Weiss ◽  
...  
Keyword(s):  
Power System ◽  
Failure Time ◽  
Late Phase ◽  
Electric Power System ◽  
Physical Models ◽  
Severe Accident ◽  
Feed Water ◽  
The Core ◽  
Lower Head ◽  
Station Blackout

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

scholarly journals Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

2013 ◽  
Vol 2013 ◽  
pp. 1-15 ◽  
Author(s):  
Andrej Prošek ◽  
Leon Cizelj
Keyword(s):  
Management Strategies ◽  
Severe Accident ◽  
Stress Tests ◽  
Pressurized Water ◽  
Makeup Water ◽  
Reactor Coolant ◽  
The Core ◽  
Station Blackout ◽  
Accident Sequences

Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

A Method for Online Calibration of MAAP Simulations in a Severe Accident Management System Database

2017 ◽  
Author(s):  
Weifeng Xu ◽  
Fangqing Yang ◽  
Peng Chen ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Outlet Temperature ◽  
Online Calibration ◽  
Support Plate ◽  
Initial Event ◽  
Accident Management ◽  
Accident Simulation ◽  
Plate Failure

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.

scholarly journals Analysis of criticality of melt during severe accidents in reactor vessel

2018 ◽  
pp. 3-10
Author(s):  
Yu. Kovbasenko ◽  
Yevgen Bilodid
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Cooling Water ◽  
Nuclear Fission ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Emergency Situations ◽  
The Core

The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

scholarly journals Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

2020 ◽  
Vol 8 ◽  
Author(s):  
Hyoung Tae Kim ◽  
Jin Ho Song ◽  
Rae-Joon Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Hydrogen Generation ◽  
Alternate Current ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Active Safety Systems ◽  
Station Blackout ◽  
Safety Systems

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

Sign in / Sign up

Export Citation Format

Share Document