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.

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.

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.

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.

Development of Best-Estimate ECCS Evaluation Methodology for APR1400

2009 ◽  
Author(s):  
Seok-Ho Lee ◽  
Mun-Soo Kim ◽  
Han-Gon Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Evaluation Methodology ◽  
Preliminary Calculation ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
Best Estimate ◽  
Design Characteristics ◽  
Core Cooling

Advanced Power Reactor 1400 (APR1400) is an evolutionary Pressurized Water Reactor (PWR) equipped with such advanced features as the Direct Vessel Injection (DVI), the Fluidic Device (FD) in the Safety Injection Tank (SIT), and the In-containment Refueling Water Storage Tank (IRWST) in the Emergency Core Cooling System (ECCS). To verify the performance of these advanced features, more realistic performance evaluation methodology is desired since existing methodologies use too conservative assumptions which cause negative biases to these features. In this study, therefore, a best estimate evaluation methodology for the APR1400 ECCS under large break loss of cooling accident (LBLOCA) is developed targeting operating license of the Shin Kori 3&4 nuclear power plants (SKN 3&4), the first commercial APR1400 plants. On this purpose, a variety of existing best estimate evaluation methodologies previously used are reviewed. As a result of this review, a methodology named KREM is selected for this study. The KREM is based on RELAP5/MOD3.1K and has been used for Korean operating plants since 2002 when it was first approved by Korean regulation. For this study, RELAP5/MOD3.3 (Patch 3), the latest version of RELAP series is selected since it could appropriately simulate the multi-dimensional phenomena for the APR1400 design characteristics. To quantify the code accuracy, analyses covering experimental data have been performed for 36 kinds of separated effect tests (SETs) and integral effect tests (IETs). The uncertainty in the peak cladding temperature (PCT) of the APR1400 is evaluated preliminarily. Based on the preliminary calculation, final uncertainty quantification and bias evaluation are performed to obtain the licensing PCT for Shin-Kori 3&4 plants and the result shows that the LBLOCA licensing acceptance criteria are well satisfied.

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.

CANDU 600 Severe Accident Assessment

2008 ◽  
Author(s):  
Gheorghe Negut ◽  
Ilie Prisecaru ◽  
Alexandru Catana ◽  
Daniel Dupleac
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Pressure Tube ◽  
Severe Accident ◽  
Nuclear Power Reactor ◽  
Severe Accidents ◽  
Start Up ◽  
Reactor Geometry ◽  
Simultaneous Loss ◽  
Core Cooling

Romania is now a UE member since January first 2007. New challenges are for our country that includes, also, their nuclear power reactors. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU are operated PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model severe accidents for these types of reactors. Starting from previous studies a CANDU degraded core thermalhydraulic model was developed. The initiating event is a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

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.

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