Post Small-Break LOCA Long Term Core Cooling Performance in ACME Test Facility

2016 ◽  
Author(s):  
Zhanfei Qi ◽  
Sheng Zhu
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Margin ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Features ◽  
Core Cooling

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.

Correction of Zero-Drift of Differential Pressure Flow Meters of ACME Facility

2014 ◽  
Author(s):  
Mingtao Cui ◽  
Tao Zhang
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Cooling System ◽  
Differential Pressure ◽  
Zero Drift ◽  
Test Facility ◽  
Passive Safety ◽  
Pressure Flow ◽  
Flow Meters ◽  
Core Cooling

ACME facility (Advanced Core-cooling Mechanism Experiment) is a large-scale test facility used to validate the performance of passive core-cooling system under SBLOCA (Small Break Lost of Coolant Accident) for the CAP1400, an upgraded passive safety nuclear power plant of AP1000. To simulate the features of passive safety system properly, DELTABAR, a kind of differential pressure flow meter, is designed to measure different mass flow of ACME. Because of the low pressure loss of DELTABAR, Zero-Drift problem of differential pressure flow meters in ACME is amplified, and some of the measured values are distorted seriously. To minimize the influence of Zero-Drift, analysis on zero-drift phenomenon is made, and a compensation method is proposed. The method is applying to PBL flow meters, and the result shows that the method is applicable.

Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility

2016 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
System Design ◽  
Nuclear Power ◽  
Reactor Core ◽  
Safety System ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water ◽  
Power Simulation ◽  
Passive Safety System ◽  
Core Cooling

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.

Containment Sump Active Strainers

2005 ◽  
Author(s):  
A. R. Mehta ◽  
A. J. Bilanin ◽  
J. Hamel ◽  
A. Kaufman
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Source ◽  
Nuclear Regulatory Commission ◽  
Test Facility ◽  
Plant Design ◽  
Test Environment ◽  
Pressurized Water

The containment sump, also known as emergency or recirculation sump, is part of the Emergency Core Cooling System (ECCS). Every nuclear power plant is required by regulations to have an ECCS to mitigate a design basis accident. The containment sump of a Pressurized Water Reactor (PWR) collects reactor coolant and chemically reactive spray solutions following a Loss of Coolant Accident (LOCA). The containment sump serves as the water source to support long-term recirculation. This water source, the related pump inlets and the piping between the source and inlets are all important safety components. Suppression pools in Boiling Water Reactors (BWRs) serve the same purpose as PWR containment sumps. Historically, a passive debris screen has been used to prevent debris from entering the ECCS suction lines surrounding the containment sump. Previous incidents demonstrated that the potential for excessive head loss across the containment sump screens exists because of the accumulation of debris on the containment sump. Because of this, the US Nuclear Regulatory Commission (NRC) has concluded that containment sump blockage is a potential concern for PWRs. US BWRs were required to conduct plant-specific evaluations of their suction strainer performance and, as required, modify their plant design. While all US PWRs are required to resolve this Generic Safety Issue (GSI-191), containment sump blockage continues to be a major concern for both BWRs and PWRs internationally. This paper describes the GE Active Strainer design, one of several strainers developed to resolve this generic safety issue. The Active Strainer presents an innovative and novel method of addressing containment sump blockage. This strainer employs a rotating, or “active”, plow and brush that sweep over a perforated surface. By keeping the perforated surface free of debris, fluid is allowed to pass through, providing sufficient coolant to the ECCS pumps to support long-term recirculation. Due to the unique method by which the Active Strainer filters coolant, a test program was developed to demonstrate its functionality and viability. Intrinsic differences between passive and active solutions make previous methods of testing obsolete for the GE Active Strainer. Moreover, the complex and varying geometries and conditions of actual plant containment sumps are difficult to replicate. Therefore, a methodology was developed to ensure prototypical test environment and strainer debris loads in a scaled test facility. This paper will discuss the GE Active Strainer design, the testing conducted and subsequent conclusions.

Analysis of Small-Break LOCA at ACME Test Facility Using RELAP5/MOD3

2016 ◽  
Author(s):  
Xu Caihong ◽  
Shi Guobao ◽  
Fan Pu
Keyword(s):  
Test Data ◽  
Cooling System ◽  
Test Facility ◽  
Test Cases ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Cooling Mechanism ◽  
Core Cooling

The Advanced Core-cooling Mechanism Experiment (ACME) is conducted to investigate the performance of passive core-cooling system (PXS) for the advanced CAP1400 Pressurized Water Reactor (PWR). The small-break LOCA experiments conducted at ACME integrated test facility are simulated with a SNERDI modified version of RE-LAP5/MOD3 code. Several typical SBLOCA test cases are simulated and one case (2 inch cold leg break) is presented in this paper. And the predicted results are compared with the test data to assess the performance of the modified code. The calculated results agree reasonably well with the test data.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

scholarly journals Research and Evaluation for Passive Safety System in Low Pressure Reactor

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
Peng Chuanxin ◽  
Zhuo Wenbin ◽  
Chen Bingde ◽  
Nie Changhua ◽  
Huang Yanping
Keyword(s):  
Steady State ◽  
Satisfactory Agreement ◽  
Nuclear Power ◽  
Low Pressure ◽  
Safety System ◽  
Test Facility ◽  
Passive Safety ◽  
Passive Safety System ◽  
Calculation Results ◽  
Core Cooling

Low pressure reactor is a small size advanced reactor with power of 180 MWt, which is under development at Nuclear Power Institute of China. In order to assess the ability and feasibility of passive safety system, several tests have been implemented on the passive safety system (PSS) test facility. During the LOCA and SBO accident, the adequate core cooling is provided by the performance of passive safety system. In addition the best-estimate thermal hydraulic code, CATHARE V2.1, has been assessed against cold leg LOCA test. The calculation results show that CATHARE is in a satisfactory agreement with the test for the steady state and transient test.

Emergency Operating Procedures (EOPs) for the AP1000 Simulator

2006 ◽  
Author(s):  
Yuichi Hayashi ◽  
Gianfranco Saiu ◽  
Richard F. Wright
Keyword(s):  
Pressure Control ◽  
Passive Safety ◽  
Background Information ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Investment Protection ◽  
Plant Safety ◽  
Safety Features ◽  
Core Cooling ◽  
Emergency Operating

The AP1000 is two-loop 1100 MWe advanced pressurized water reactor (PWR) that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 30 years of operating PWR experience. The AP1000 final design certification was approved by the NRC in December, 2005. A total of 34 Emergency Operating Procedures (EOPs) for operation of the AP1000 simulator have been prepared based on the AP1000 Emergency Response Guidelines (ERGs), background information documents and detailed plant information. These include 28 EOPs at power and 6 EOPs during shutdown. The AP1000 ERGs were developed by using the generic ERGs for the low pressure reference PWR plant as a basis. The AP1000 design differences from the reference plant were reviewed and reflected in the process of developing operational steps in each ERG. The provisions of the AP1000 PRA were also reviewed and incorporated into the ERGs. Although the AP1000 design does not require operator actions for the first 72 hours after accidents, the operator actions with both safety-related and nonsafety-related equipment have an important role to mitigate the consequence of accidents. In the event of a steam generator tube rupture (SGTR), although the AP1000 is designed so that no operator actions are required to recover from the event, there are actions that can be taken by the operator to limit the release of radioactive effluents from the ruptured SG. These actions include isolation of the ruptured SG and depressurization of the reactor coolant system (RCS) to terminate primary-to-secondary leakage, restoring reactor coolant inventory to ensure adequate core cooling and plant pressure control. It is expected that these operator actions should be incorporated into the ERG to reduce the fission product release. To support the development of the AP1000 ERGs, several transient and accident analyses were performed. These include analyses for LOCA, post-LOCA cooldown and depressurization, passive safety system termination, SGTR and faulted SG isolation. These analyses results were incorporated into the ERG background information documents. In the event of SGTR, several cases were analyzed, including consideration of operator recovery actions. These cases were modeled using the best-estimate state-of-art RELAP5 code. The analyses results show that operator recovery actions are effective for SGTR to be placed under operator control.

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