Modeling of Thermal Hydraulics Aspects of Combined Top and Bottom Water Reflood Experiment PARAMETER-SF2 Using SOCRAT 2.1 Code

2008 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Flow Rate ◽  
Nuclear Power ◽  
Cooling System ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Vver Fuel ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The PARAMETER-SF2 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in April 3, 2007 and was the second of two experiments to be performed in the frame of ISTC 3194 Project. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. After the maximum cladding temperature of 1750K was reached in the bundle during PARAMETER-SF2 test, the top flooding (flow rate 40g/s) was begun and later approximately in 30 s the bottom flooding (flow rate 100g/s) was initiated. Two-phase (water and steam) flow determined the fuel assembly cooling conditions. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT 2.1 was used for the calculation of PARAMETER-SF2 experiment. Thermal hydraulics in PARAMETER-SF2 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT 2.1 were compared with experimental data concerning different aspects of thermal hydraulics behavior including convective and radiative heat transfer in the bundle and the CCFL (counter-current flooding limitation) phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF2 test.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Posttest Calculations of PARAMETER-SF4 Experiment With Air Ingress and Water Reflood Using SOCRAT/V3 Computer Modeling Code

2011 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
Experimental Data ◽  
Computer Modeling ◽  
Nuclear Power ◽  
Cooling System ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Vver Fuel ◽  
Design Basis ◽  
Air Ingress ◽  
Modeling Code

The PARAMETER-SF4 test conditions simulated a severe LOCA (Loss of Coolant Accident) NPP (nuclear power plant) sequence in which the overheated up to 1700–2300K core would be reflooded from the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in July 21, 2009, and was the fourth of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH” (scientific and industrial association LUTCH), Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators. Heating was carried out electrically using tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVER (water-water energetic reactor, Russian type of pressurized water reactor). After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF4 test, the bottom flooding was initiated. The important feature of PARAMETER-SF4 test was the air ingress phase during which the air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in a different way in comparison to oxidation in the steam. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the calculation of PARAMETER-SF4 experiment. Thermal hydraulics in PARAMETER-SF4 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V3 were compared with experimental data concerning different aspects of air ingress phase and thermal hydraulics behavior during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF4 test.

scholarly journals CFD Analysis of Passive Autocatalytic Recombiner

2011 ◽  
Vol 2011 ◽  
pp. 1-9
Author(s):  
B. Gera ◽  
P. K. Sharma ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Maximum Temperature ◽  
Convection Flow ◽  
Heat Of Reaction ◽  
Loss Of Coolant Accident ◽  
Flow Through ◽  
Core Cooling ◽  
Catalytic Recombination ◽  
Steam Heat

In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction produces natural convection flow through the enclosure and promotes mixing in the containment. For the assessment of the PAR performance in terms of maximum temperature of catalyst surface and outlet hydrogen concentration an in-house 3D CFD model has been developed. The code has been used to study the mechanism of catalytic recombination and has been tested for two literature-quoted experiments.

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

Use of LBB Methodology to Support the Transition Break Size Concept

2011 ◽  
Author(s):  
James E. Nestell ◽  
David W. Rackiewicz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Reactor Coolant ◽  
Design Basis ◽  
Piping Systems ◽  
Pipe Break ◽  
Core Cooling ◽  
Pipe Size

The design basis for a loss of coolant accident in nuclear power plants has previously been based on the assumption that the largest size coolant pipe instantaneously undergoes a double ended “guillotine” break (DEGB) and the resulting loss of water must be mitigated by an emergency core cooling system (ECCS) to maintain core cooling after shutdown. The U.S. Nuclear Regulatory Commission (NRC) is close to allowing a risk-informed design basis break size, called the Transition Break Size (TBS), to be used for LOCA break size assumptions for ECCS design. The TBS approach will require full safety redundancy for an ECCS system sized to handle a break of the next largest reactor coolant pipe size (rather than the largest reactor coolant pipe size), and it will allow relaxed system redundancy requirements for handling the largest pipe break size. The TBS will thereby reduce the cost of the safety-grade ECCS system in new plant designs and will increase operational flexibility in existing plants. The TBS approach is based on the results of NRC elicitation studies with piping experts regarding historical pipe performance and risk of sudden failure. The approach is non-deterministic and is a conceptual change from the largest-pipe-size break assumption. The conceptual discontinuity between deterministic and elicitation-based break size assumptions could be uncomfortable for those schooled in strictly deterministic accident analyses. In this paper we explore the “leak-before-break” (LBB) methodology as it applies to large pipe break analyses in nuclear piping systems, and show through examples that the elicitation-based TBS approach is indeed conservative when TBS results are compared with deterministic LBB evaluations of similar piping systems. Thus, LBB provides a deterministic means for showing defense in depth against LOCAs greater than the TBS break size.

Assessment of APR-1400 Emergency Core Cooling System Performance for Design Basis LOCA Redefinition

2014 ◽  
Author(s):  
Dong Gu Kang ◽  
Seung-Hoon Ahn ◽  
Soon Heung Chang ◽  
Byung-Gil Huh ◽  
Young-Seok Bang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Cooling System ◽  
Exceedance Probability ◽  
Plant Design ◽  
Diesel Generator ◽  
Start Time ◽  
Design Modification ◽  
Design Basis ◽  
Core Cooling

As a part of the efforts to develop the risk-informed regulation, alternative rulemaking of 10CFR50.46 is underway. In the rule, USNRC divided the current spectrum of LOCA break sizes into two regions, by determining a transition break size (TBS), and the LOCAs for any breaks larger than TBS would be regarded as beyond design basis accident (BDBA). A combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of BDBAs. The performance of the APR-1400 emergency core cooling system (ECCS) performance was assessed against large break LOCA applying CDPP. It was confirmed that current APR-1400 ECCS design has capability to mitigate BDB LOCA by analyzing ECCS cooling performance for BDB LOCA. The proposed CDPP was also applied to design changes of the emergency diesel generator (EDG) start time extension and power uprates with simplified assumption that the probabilistic safety assessment (PSA) data are still valid. By assumptions and considerations, the CDPP to assess ECCS performance for plant design modification was reduced to calculating conditional exceedance probability (CEP) of one sequence and comparing allowable value. The allowable CEP was used to determine whether the design change is acceptable or not, and discussions were made for acceptable nuclear power plant changes.

Assessment of Reliability of Emergency Core Cooling System (ECCS) of Bushehr Nuclear Power Plant

2009 ◽  
Author(s):  
Mohammad Sotoudeh ◽  
Kamran Sepanloo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Computer Code ◽  
Probabilistic Methods ◽  
Event Tree ◽  
Loss Of Coolant Accident ◽  
Fuel Storage ◽  
Core Cooling

Bushehr Nuclear Power Plant (BNPP) originally designed by German KWU and then modified by Russian companies is approaching commissioning. In this paper, the reliability of Emergency Core Cooling System (ECCS) of the new design is compared with the older one using both deterministic and probabilistic methods. The reliability of both systems is calculated in case of occurrence of Large Loss of Coolant Accident (Large LOCA). The new Russian design is based on the original KWU Convoy design and the applied modifications have improved its reliability. To perform the reliability analysis the event tree/fault tree method is used and the calculations are done by IAEA computer code PSAPACK. The results show that the applied modifications such as, increase of redundancy from 4 × 50% to 4 × 100%, change of cool down route, fuel storage pool cooling in all four trains have reduced the ECCS unavailability from 5.9 E−4 to 1.7 E−4. Furthermore, based on the results it is shown that both designs comply well with the IAEA recommendations on Probabilistic Safety Criteria (PSC) and INSAG-3 requirements.

External Hazard Coinciding With Small Break LOCA—Thermohydraulic Calculation With System Code ATHLET

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Kai Kosowski ◽  
Marcus Seidl
Keyword(s):  
Diesel Engines ◽  
Nuclear Power ◽  
Heat Removal ◽  
Primary Circuit ◽  
External Event ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
License Application ◽  
Thermohydraulic Calculation ◽  
Core Cooling

Abstract The safety behaviors of a nuclear power plant (NPP) after an external hazard-initiated event, as well as after a small break (SB) loss of coolant accident (LOCA), are already well known as part of the analyses made for standard license application. The coincidence of both events leads to a beyond-design basis consideration. Such a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET. The scenario assumes an external event with a LOCA caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. Furthermore, in the context of both robustness and enveloping analyses, both a loss of offsite power (LOOP) and an unavailability of the emergency diesel power supply are postulated. The NPP in the scenario considered only has access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (second quartet of emergency diesel engines), which are housed in the bunkered emergency feed building. The dedicated type of external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered. Apart from the model's assumptions, the accident sequence is explained in detail. The remaining systems for emergency core cooling are capable of handling the LOCA under such demanding boundary conditions. Long-term cooling can be ensured. Furthermore, heat removal out of the core is always sufficient. Eventually, all safety protection objectives have been complied for this beyond-design basis scenario.

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Alexander Vasiliev ◽  
Juri Stuckert
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Post Test ◽  
Core Cooling ◽  
Modeling Code ◽  
Code Quality

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

Sign in / Sign up

Export Citation Format

Share Document