Wall Materials Effects on Sheltered Indoor Doses From an SMR Hypothetical Accident Release

2021 ◽  
Author(s):  
Yamato Sugitatsu ◽  
Shripad T. Revankar
Keyword(s):  
Emergency Planning ◽  
Severe Accident ◽  
Air Concentration ◽  
Loss Of Coolant Accident ◽  
Energy Needs ◽  
Small Modular Reactor ◽  
Station Blackout ◽  
Wall Materials ◽  
Set Up ◽  
Radiological Dose

Abstract Small modular reactors (SMRs) are expected as a suitable candidate to fulfill energy needs in the future. The regulation of the emergency planning zone (EPZ) has been a controversial issue. The possibility of smaller EPZs because of their small core size and passive safety functions has still under discussion. The major emergency responses to radiological incidents in the early phase are evacuation from the area and sheltering-in-place within a building. Comparison between the dose incurred during evacuation and that with sheltering-in-place is necessary to consider the proper protective actions. This study focuses on effect of wall materials on indoor doses for sheltered population from small modular reactor severe accident. The source term came from loss of coolant accident or station blackout, and the time change of air concentration and the ground deposition data was calculated with RASCAL, a software developed by NRC to provide dose projection around the plant. Then general one-story and two-story houses were set up, and 6 wall materials were selected for calculating indoor doses. Cloudshine and groundshine were calculated with Monte Carlo methods, and the shielding function of each house was evaluated by comparing the indoor dose with outdoor dose. The result will be a basis for calculating the radiological dose for sheltered cases in case of nuclear emergency for SMR, which will be valuable to have a more effective emergency planning.

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.

scholarly journals Analysis of Processes in the Containment Using ATHLET-CD and COCOSYS Codes

2020 ◽  
pp. 27-37
Author(s):  
M. Vyshemirskyi ◽  
V. Pustovit ◽  
V. Kravchenko ◽  
D. Donskyi
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Severe Accident ◽  
Primary System ◽  
Coupled Calculation ◽  
Code Model ◽  
Loss Of Coolant Accident ◽  
Calculation Results ◽  
Station Blackout ◽  
Containment Pressure

A brief description of performed input deck modifications and results of stand-alone and coupled calculations of Dn 200 mm loss of coolant accident with simultaneous total station blackout accident scenario for Rivne Nuclear Power Plant Unit 1 (WWER‑440/V-213) with application of ATHLET-CD 3.1A and COCOSYS 2.4 codes are presented in the paper. ATHLET-CD stand-alone calculation was performed with constant containment pressure (a time dependent volume with constant pressure and temperature was used as a boundary volume for leakage). Further, mass and energy release and fission products from the primary system obtained during ATHLET‑CD stand-alone calculation were used to perform COCOSYS stand-alone calculation. In addition, coupled ATHLET-CD and COCOSYS calculation was performed. All the computer analyzes were performed until the lower head failure. ATHLET‑CD model was extended with core degradation module (ECORE), which allowed calculation of scenario until reactor pressure vessel failure. According to the results of comparative analysis, nearly the same behavior of the main parameters in the stand-alone and coupled calculation at an early phase of scenario was obtained. Some small differences occur due to distinction in behavior of water and steam mass flows released through the break and due to existence of heat transfer from the primary system structures to the containment compartments during coupled calculation of transient. As for middle and late phases of the accident, some differences between stand-alone and coupled calculation results for analyzed scenario are present. These differences are caused by different total fission products and aerosols release from the reactor coolant system to the containment compartments. The above information allows recommending application of coupled code/model versions for performing the computer severe accident analyses.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

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.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Severe accident management measures for a generic German PWR. Part I: Station blackout

2018 ◽  
Vol 122 ◽  
pp. 217-228 ◽  
Author(s):  
P. Wilhelm ◽  
M. Jobst ◽  
Y. Kozmenkov ◽  
F. Schäfer ◽  
S. Kliem
Keyword(s):  
Severe Accident ◽  
Management Measures ◽  
Accident Management ◽  
Station Blackout

The Severe Accident Analysis of CPR1000 Based on MELCOR Code

2013 ◽  
Author(s):  
Longze Li ◽  
Mingjun Wang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Failure Time ◽  
Water Pressure ◽  
Severe Accident ◽  
Feed Water ◽  
Calculation Results ◽  
Station Blackout ◽  
Accident Response ◽  
Basic Standard

The severe accident of CPR1000 caused by station blackout with the SG safety valve failure is simulated and analyzed using MELCOR code in this work. The CPR1000 power plant severe accident response process and the results with three different assumptions, which are no the seal leakage nor the auxiliary feed water, the seal leakage and auxiliary feed water exist, the seal leakage exist but no auxiliary feed water separately, are analyzed. According to the calculation results, without the seal leakage and auxiliary feed water, pressure vessel would fail at 9576 s. When auxiliary feed water was supplied, pressure vessel’s failure time would delay nearly 30000s. When the seal leakage exists, pressure vessel’s failure time would delay about 50 s. The results are meaningful and significant for comprehending the detailed process of severe accident for CPR1000 nuclear power plant, which is the basic standard for establishing the severe accident management guideline.

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

2016 ◽  
Vol 86 ◽  
pp. 87-96 ◽  
Author(s):  
Shasha Yin ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G.H. Su ◽  
...  
Keyword(s):  
Severe Accident ◽  
Accident Scenario ◽  
Small Modular Reactor

Severe Accident Analysis for BWR With Containment Outer Pool

2012 ◽  
Author(s):  
Koki Yoshimura ◽  
Kohei Hisamochi
Keyword(s):  
Cooling System ◽  
Water Injection ◽  
Heat Removal ◽  
Severe Accident ◽  
Gas Temperature ◽  
Passive System ◽  
Active Components ◽  
Active System ◽  
Station Blackout ◽  
Materials Used

Newly designed plants, e.g., next-generation light water reactor or ESBWR, employ a passive containment cooling system and have an enhanced safety with RHRs (Residual Heat Removal system) including active components. Passive containment cooling systems have the advantage of a simple mechanism, while materials used for the systems are too large to employ these systems to existing plants. Combination of passive system and active system is considered to decrease amount of material for existing plants. In this study, alternatives of applying containment outer pool as a passive system have been developed for existing BWRs, and effects of outer pool on BDBA (Beyond Design Basis Accident) have been evaluated. For the evaluation of containment outer pool, it is assumed that there would be no on-site power at the loss of off-site power event, so called “SBO (Station BlackOut)”. Then, the core of this plant would be uncovered, heated up, and damaged. Finally, the reactor pressure vessel would be breached. Containment gas temperature reached the containment failure temperature criteria without water injection. With water injection, containment pressure reached the failure pressure criteria. With this situation, using outer pool is one of the candidates to mitigate the accident. Several case studies for the outer pool have been carried out considering several parts of containment surface area, which are PCV (Pressure Containment vessel) head, W/W (Wet Well), and PCV shell. As a result of these studies, the characteristics of each containment outer pool strategies have become clear. Cooling PCV head can protect it from over-temperature, although its effect is limited and W/W venting can not be delayed. Cooling suppression pool has an effect of pressure suppressing effect when RPV is intact. Cooling PCV shell has both effect of decreasing gas temperature and suppressing pressure.

Sign in / Sign up

Export Citation Format

Share Document