Severe accident management measures for a generic German PWR. Part II: Small-break loss-of-coolant accident

2018 ◽  
Vol 122 ◽  
pp. 280-296 ◽  
Author(s):  
M. Jobst ◽  
P. Wilhelm ◽  
Y. Kozmenkov ◽  
S. Kliem
Keyword(s):  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Management Measures ◽  
Accident Management ◽  
Loss Of Coolant

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.

scholarly journals Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed

2011 ◽  
Vol 175 (3) ◽  
pp. 572-593 ◽  
Author(s):  
César Queral ◽  
Juan González-Cadelo ◽  
Gonzalo Jimenez ◽  
Ernesto Villalba
Keyword(s):  
High Pressure ◽  
Management Actions ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant

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

Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

2002 ◽  
Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  
Keyword(s):  
Flow Model ◽  
Correction Term ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Primary System ◽  
Two Phase ◽  
Flow Rates ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

scholarly journals SUBSTANTIATION OF MODERNIZED BLACKOUT & LOSS-OF-COOLANT ACCIDENT MANAGEMENT STRATEGY AT NUCLEAR POWER PLANTS WITH WWER

2020 ◽  
Vol 2 (61) ◽  
pp. 70-77
Author(s):  
V. Skalozubov ◽  
◽  
V. Spinov ◽  
D. Spinov ◽  
Т. Gablaya ◽  
...  
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Management Strategies ◽  
Computational Modelling ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Safety Systems

The analysis of the known results of RELAP5/V.3.2 simulation for loss of coolant & blackout accidents at WWER nuclear power plants showed that the design accident management strategies with design passive safety systems do not provide the necessary safety conditions for the maximum permissible temperature of fuel claddings, the minimum permissible level of coolant in the reactor and feed water in the steam generators. A conservative thermohydrodynamic model for a design and modernized blackout & loss-of-coolant accident management strategy at a nuclear power plant with WWER has been developed. Design passive safety systems carry out the design accident management strategy: pressurizer safety valves, secondary steam relief valves, and hydraulic reservoirs of the emergency core cooling system of the reactor. Promising afterheat removal passive systems and the reactor level and steam generator water level control systems carry out the modernized blackout & loss-of-coolant accident management strategy. The main conservative assumptions of the presented model of blackout & loss-of-coolant accidents: complete long-term failure of all electric pumps of active safety systems, the temperature of nuclear fuel in the central part of the fuel matrix is assumed as the maximum allowable one, effect of “run down” flow of a turbine feed pump and the coolant level in pressurizer on accident process is not considered. Computational modelling has found that violations of the safety conditions are over the entire range of leak sizes for the design blackout & loss-of-coolant accident management strategy. For the modernized blackout & loss-of-coolant accident management strategy, safety conditions are provided for 72 hours of the accident and more. The presented results of computational modelling of blackout accident management strategies for nuclear power plants can be used to modernize and improve symptom-informed emergency instructions and guidelines for the severe accident management at nuclear power plants with WWER. Application of the results of computational modelling of blackout accident management strategies is generally not substantiated for other types of reactor facilities. In this case, it is necessary to develop calculated models for blackout accident management taking into account the specifics of the structural and technical characteristics and operating conditions for safety related systems of nuclear power plants.

Problems With Cooling of Overheated Fuel Channels in RBMK-Type Reactors

2008 ◽  
Author(s):  
Eugenijus Uspuras ◽  
Algirdas Kaliatka
Keyword(s):  
Heat Removal ◽  
Water Source ◽  
Severe Accident ◽  
Management Guidelines ◽  
Management Measures ◽  
Accident Management ◽  
Feed Strategy ◽  
Significant Probability ◽  
Reactor Types

One of the most dangerous beyond design basis accidents for all types of nuclear reactors is the loss of long-term heat removal from the core. In RBMK-type reactors, this initiating event, which can lead to the worst consequences, has significant probability to occur in comparison to other type of BDBA. The most effective accident mitigation measure in this case is “bleed and feed” strategy — similar as is recommended for other light water reactor types. In this paper the challenges, which are meet in case of cooling of overheated fuel channels in RBMK-type reactors, are discussed. The simulation results of BDBA using RELAP5/MOD3.3 code are presented. Accident management measures (de-pressurization of reactor cooling circuit and injection of water from non-regular water source) are evaluated in respect of dangerous pressure increase and thermal shock in fuel channels. These results were used during development of severe accident management guidelines for RBMK-1500 at Ignalina NPP.

The Development and Application of SCDAP-3D©

2002 ◽  
Author(s):  
E. W. Coryell ◽  
E. A. Harvego ◽  
L. J. Siefken
Keyword(s):  
Computer Code ◽  
Alternative Fuel ◽  
Severe Accident ◽  
Initial Development ◽  
Standard Problem ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Environmental Laboratory ◽  
Reactor Accidents ◽  
Thermal Hydraulic Analysis

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.

scholarly journals ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

2018 ◽  
Vol 2018 ◽  
pp. 1-24
Author(s):  
J. C. de la Rosa Blul ◽  
S. Brumm ◽  
F. Mascari ◽  
S. J. Lee ◽  
L. Carenini
Keyword(s):  
Failure Time ◽  
Hydrogen Generation ◽  
Severe Accident ◽  
Primary System ◽  
Similar Amount ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Expansion Model ◽  
Gas Expansion ◽  
Comparison Exercise

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

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