Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds

2008 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Aladar A. Csontos ◽  
Timothy R. Lupold ◽  
Chia-Fu Sheng
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Primary Water ◽  
Water Reactors ◽  
The Impact ◽  
Growth Analyses

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.

scholarly journals Major Achievements and Prospect of the ATLAS Integral Effect Tests

2012 ◽  
Vol 2012 ◽  
pp. 1-18 ◽  
Author(s):  
Ki-Yong Choi ◽  
Yeon-Sik Kim ◽  
Chul-Hwa Song ◽  
Won-Pil Baek
Keyword(s):  
Large Scale ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Hydraulic Test ◽  
Validation Data ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Integral Effect

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

Integral Effect Test for a Inadvertent Opening of POSRV Using the ATLAS Facility

2010 ◽  
Author(s):  
Seok Cho ◽  
Ki-Yong Choi ◽  
Hyun-Sik Park ◽  
Kyoung-Ho Kang ◽  
Yeon-Sik Kim ◽  
...  
Keyword(s):  
Nuclear Industry ◽  
Test Facility ◽  
System Response ◽  
Integral Test ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Reduced Height ◽  
Test Loop ◽  
Integral Effect ◽  
Effect Test

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.

Nonlinear Zero-Dynamics Attacks Targeting Nuclear Power Plants

2020 ◽  
Author(s):  
Jacob A. Farber ◽  
Daniel G. Cole
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactors ◽  
Zero Dynamics ◽  
Pressurized Water ◽  
Undesirable State ◽  
System Pressure ◽  
Internal States ◽  
Water Reactors ◽  
Critical Subsystem

Abstract A zero-dynamics attack allows an attacker to input some control action that results in zero measurable output but nonzero response of the internal states. This paper extends previous works on zero-dynamics attacks to nonlinear system dynamics. This is accomplished using invariant subspace techniques that identify the subspace on which zero dynamics exist. An iterative algorithm is presented to identify both this subspace and the resulting zero dynamics of the system. These methods are implemented on a model of a pressurizer in a nuclear power plant, which is a critical subsystem of pressurized water reactors that monitors and controls the system pressure and coolant inventory. This implementation is done by analyzing all combinations of attackable signals. These attackable signals are the set of all system inputs and outputs. From this analysis, there are eight unique combinations of attacked actuators and sensors that result in zero-dynamics attacks. These combinations are characterized by stability and damage time, where damage time is the time it takes to reach some undesirable state. The damage times range from half a day to sixteen days, depending on the number of signals the attacker has access to. These results demonstrate that the physics of the pressurizer system creates some vulnerabilities to zero-dynamics attacks. This work provides plant designers with tools to identify which subsystems are most susceptible to zero-dynamics attacks and might require additional defenses.

The Research and Comparison of Thermal-Hydraulic Test Facilities for Pressurized Water Reactors (PWRs)

2017 ◽  
Author(s):  
Yuan Lu ◽  
Changzhi Xiao
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Nuclear Safety ◽  
Test Facility ◽  
Scaling Analysis ◽  
Volume Data ◽  
Integral Test ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
Wide Range

Recently, nuclear power safety draws more attention after Fukushima nuclear accident, for which it is essential to construct a large number of test facilities simulating possible issues occurred in the reactor. The thermal-hydraulic test facility is extensively used to simulate thermal-hydraulic response during a loss of coolant accident (LOCA) or an operational transient which can minimize the nuclear safety accidents. This paper focus on the research of thermal-hydraulic test facilities of PWRs in different countries. All of facilities were designed by scaling analysis method. Meanwhile, a wide range of data comprising of power data, pressure data, volume data, configuration and a series of nuclear safety test data is compared in over ten test facilities. Based on above comparable data and relevant research, the main conclusions are as follows: Scaling analysis lays a solid foundation for the design and construction of scale-down nuclear reactor thermal-hydraulic test facilities. This would provide the reference for choosing scaling concepts in the reactor integral test facility.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

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