ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

2006 ◽  
Author(s):  
N. Reinke ◽  
K. Neu ◽  
H.-J. Allelein
Keyword(s):  
Fission Products ◽  
Feasibility Studies ◽  
Current Status ◽  
Light Water ◽  
Fast Running ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Station Blackout ◽  
Water Reactors ◽  
Reactor Types

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors

2011 ◽  
Author(s):  
Keisuke Okumura ◽  
Shiho Asai ◽  
Yukiko Hanzawa ◽  
Tsutomu Okamoto ◽  
Hideya Suzuki ◽  
...  
Keyword(s):  
Parametric Study ◽  
Computational Study ◽  
Estimation Method ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Light Water Reactors ◽  
Nuclear Fuels ◽  
Light Water ◽  
Water Reactors ◽  
High Level

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

2006 ◽  
Author(s):  
Gert Sdouz
Keyword(s):  
Source Term ◽  
Management Strategies ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Management Measures ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Accident Sequences ◽  
Comprehensive Study

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the untightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the “Station Blackout”-sequence and the “Large Break LOCA”. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a “Large Break LOCA”-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a “Station Blackout”-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors.

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.

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