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

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.

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

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance ◽

10.1115/icone17-75516 ◽

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.

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

Volume 9: Heat Transfer, Fluid Flows, and Thermal Systems, Parts A, B and C ◽

10.1115/imece2009-10960 ◽

2009 ◽

Cited By ~ 1

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.

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Development and Application of CAP1400 Probabilistic Safety Assessment

Volume 4: Nuclear Safety, Security, Non-Proliferation and Cyber Security; Risk Management ◽

10.1115/icone25-66651 ◽

2017 ◽

Author(s):

Qinfang Zhang ◽

Wenhui Zhan ◽

Yongping Qiu ◽

Zhaohua Li ◽

Jiandong He

Keyword(s):

Safety Assessment ◽

Cooling System ◽

Detection System ◽

Fire Detection ◽

Design Modification ◽

Design Change ◽

Probabilistic Safety Assessment ◽

Core Cooling ◽

And Control ◽

Probabilistic Safety

A full scope Probabilistic Safety Assessment (PSA) is developed for CAP1400 and some design modification was performed based on the CAP1400 PSA insights. In this paper, a brief introduction of CAP1400 PSA is presented, including analysis method, procedure, results and insights obtained and design improvements. The CAP1400 PSA includes internal events level 1 PSA, low power and shutdown PSA, level 2 PSA, level 3 PSA, internal fire PSA, internal flooding PSA, seismic margin analysis and seismic PSA, etc. The CAP1400 PSA is developed based on PSA methods acknowledged internationally and in compliance with the requirement of PSA related standards, such as ASME PSA standards and Chinese PSA standards. The design improvements derived from PSA results are also introduced in this paper, including design change of passive core cooling system (PXS), diverse design of instrument and control system, design change of in-vessel retention (IVR) of molten core debris, installation of very early warning fire detection system and design change of flooding protection of PXS valves.

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The Analysis of AP1000 Beyond Design Basis DEDVI Accident

Volume 6: Beyond Design Basis Events; Student Paper Competition ◽

10.1115/icone21-15140 ◽

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.

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Use of LBB Methodology to Support the Transition Break Size Concept

ASME 2011 Pressure Vessels and Piping Conference: Volume 1 ◽

10.1115/pvp2011-57680 ◽

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.

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

Volume 10: Heat Transfer, Fluid Flows, and Thermal Systems, Parts A, B, and C ◽

10.1115/imece2008-69043 ◽

2008 ◽

Cited By ~ 1

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.

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Revised LOCA Frequency Estimates From an Expert Elicitation Process

Volume 6: Materials and Fabrication, Parts A and B ◽

10.1115/pvp2008-61562 ◽

2008 ◽

Author(s):

Paul M. Scott ◽

Robert Lee Tregoning ◽

Lee Richard Abramson

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Cooling System ◽

Nuclear Regulatory Commission ◽

Expert Elicitation ◽

Sensitivity Analyses ◽

Piping System ◽

Plant Design ◽

Elicitation Process ◽

Core Cooling

The double-ended-guillotine break (DEGB) criterion of the largest primary piping system in the plant, which generally provides the limiting condition for the emergency core cooling system requirements, is widely recognized as an extremely unlikely event. As a result, the US Nuclear Regulatory Commission (NRC) is considering a risk-informed revision of the design-basis break size requirements for commercial nuclear power plants. In support of this effort, loss-of-coolant accident (LOCA) frequency estimates were developed using an expert elicitation process by consolidating service history data and insights from probabilistic fracture mechanics (PFM) studies with knowledge of plant design, operation, and material performance. This paper describes, and presents the results for, two of the sensitivity analyses conducted as part of this effort (overconfidence adjustment and aggregation method) to examine the assumptions, structure, and techniques used to process the elicitation responses to develop group estimates of the LOCA frequency estimates.

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

Volume 10: Heat and Mass Transport Processes, Parts A and B ◽

10.1115/imece2011-63626 ◽

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.

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Multi-Pressure Surface Condenser Performance Evaluation: A Case Study

ASME 2004 Power Conference ◽

10.1115/power2004-52089 ◽

2004 ◽

Author(s):

Komandur S. Sunder Raj

Keyword(s):

Power Plant ◽

Nuclear Power ◽

Performance Modeling ◽

Plant Design ◽

Turbine Generator ◽

Modeling Tools ◽

Pressure Surface ◽

Design Basis ◽

Predicting Performance

Surface condensers for power plant applications are generally specified and designed following turbine-condenser optimization studies. The turbine manufacturer provides turbine-generator performance data (thermal kit) at the very outset of plant design when the condenser is usually a black box and not much is known about its design. The turbine-generator guarantee would then be based on a specified condenser pressure that may or may not be attainable once the condenser is actually specified and designed. The condenser pressure used for the turbine performance guarantee might assume a single-pressure condenser while the actual design might be a multi-pressure condenser. In order to properly predict and monitor the performance and conduct diagnostics on a multi-pressure condenser, it is important to understand the design basis and develop an accurate model using performance modeling tools. The paper presents a multi-pressure condenser case study for a 600 Mwe nuclear power plant. The paper discusses the design basis used, interface between the turbine and condenser, use of a performance modeling tool for predicting performance, determining capacity losses attributable to the condenser and conducting diagnostics.

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