Assessment of ASME Code Examinations on Regenerative, Letdown and Residual Heat Removal Heat Exchangers

2005 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen ◽  
S. E. Cumblidge ◽  
G. A. Tinsley ◽  
B. Lydell ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Heat Exchangers ◽  
Power Plants ◽  
National Laboratory ◽  
Operating Experience ◽  
Heat Removal ◽  
Pacific Northwest National Laboratory ◽  
Regenerative Heat ◽  
Asme Code ◽  
Residual Heat

Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical factors, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principal reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers, regenerative heat exchangers, and residual (decay) heat removal heat exchangers. Design, operating, common preventative maintenance practices, and potential degradation mechanisms were reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp® database and covers PWR plants worldwide for the period 1970–2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers was performed. The results were then reviewed to discuss the examinations relative to plant safety and occupational radiation exposures.

Industrial Technical Development and Qualification of Highly Efficient Stainless Steel Plates and Fins Heat Exchanger for Heat Removal Supercritical CO2 Brayton Cycle Applied to Nuclear Power Plants

2020 ◽  
Author(s):  
Sarah Tioual-Demange ◽  
Gaëtan Bergin ◽  
Thierry Mazet ◽  
Luc de Camas
Keyword(s):  
Stainless Steel ◽  
Heat Exchanger ◽  
Heat Exchangers ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Eutectic Phase ◽  
Steel Plates ◽  
Brayton Cycle

Abstract The sCO2-4-NPP european project aims to develop an innovative technology based on supercritical CO2 (sCO2) for heat removal to improve the safety of current and future nuclear power plants. The heat removal from the reactor core will be achieved with multiple highly compact self-propellant, self-launching, and self-sustaining cooling system modules, powered by a sCO2 Brayton cycle. Heat exchangers are one of the key components required for advanced Brayton cycles using supercritical CO2. Fives Cryo company, a brazed plates and fins heat exchangers manufacturer, with its expertise in thermal and hydraulic design and brazing fabrication is developing compact, and highly efficient stainless steel heat exchanger solution for sCO2 power cycles, thanks to their heat exchange capability with low pinch and high available flow sections. The aim of the development of this specific heat exchanger technology is to achieve an elevated degree of regeneration. For this matter, plates and fins heat exchanger is a very interesting solution to meet the desired thermal duty with low pressure drop leading to a reduction in size and capital cost. The enhancement of the mechanical integrity of plates and fins heat exchanger equipment would lead to compete with, and even outweigh, printed circuit heat exchangers technology, classically used for sCO2 Brayton cycles. sCO2 cycle conditions expose heat exchangers to severe conditions. Base material selection is essential, and for cost reasons, it is important to keep affordable heat-resistant austenitic stainless steel grades, much cheaper than a nickel-based alloy. Another advantage of high compactness of plates and fins heat exchangers is the diminution of the amount of material used in the heat exchanger manufacturing, decreasing even more its cost. The challenge here is to qualify stainless steel plates and fins heat exchangers mechanical resistance, at cycle operating conditions, and meet with pressure vessels codes and regulations according to nuclear requirements. One critical point in the development of the heat exchangers is the design of the fins. As secondary surface, they allow the maximization of heat transfer at low pressure drop. At the same time mechanical strength has to be guaranteed. To withstand high pressure, fins thickness has to be significant, which makes the implementation complicated. Efforts were dedicated to successfully obtain an optimal shape. Forming of fins was therefore improved compared to conventional techniques. Important work was undertaken to define industrial settings to flatten the top of the fins leading to a maximum contact between the brazing alloy and the fins. Consequently brazed joints quantity is minimized inducing a diminution of the presence of eutectic phase, which is structurally brittle and limits the mechanical strength of the construction. A metallurgical study brings other elements leading to the prevention of premature rupture of the brazed structure. The idea is to determine an optimized solidification path and to identify a temperature range and holding time where the brazed joint is almost free of eutectic phase during the assembly process in the vacuum furnace.

Technical Basis for Revision of Inspection Requirements for Regenerative and Residual Heat Exchangers

2006 ◽  
Author(s):  
Warren Bamford ◽  
Bruce Bishop ◽  
Richard Haessler ◽  
Mark Bowler
Keyword(s):  
Heat Exchanger ◽  
Heat Exchangers ◽  
Ultrasonic Inspection ◽  
Regenerative Heat ◽  
Flaw Tolerance ◽  
Nuclear Plants ◽  
History Of ◽  
Technical Basis ◽  
Very High ◽  
Residual Heat

Section XI imposed volumetric inservice inspection requirements on heat exchangers in nuclear plants after most of this equipment was designed and installed. Consequently the equipment was not designed for ultrasonic examination, and in some cases such volumetric examination is not justified. The man-rem dose received from the ultrasonic inspection of some of these components is very high, and there are no known mechanisms of degradation; thus, the volumetric inspection serves no useful purpose. With the use of the newly approved code case, N706, volumetric and surface inspection of the regenerative and residual heat exchangers in PWR plants may be replaced with a visual inspection. These two heat exchangers have high irradiation fields, and both have a number of complicated weld geometries that are difficult to inspect. The regenerative heat exchanger provides preheat for the normal charging water going into the reactor coolant system (RCS). The residual heat exchanger is designed to cool the RCS during plant shut down operations. The technical basis for changing these inspection requirements was derived from four fundamental arguments: 1. The heat exchangers were carefully constructed to nuclear quality requirements. 2. They were inspected during construction, and then during service, and there is no history of degradation. 3. The flaw tolerance of the components is very high, since their duty cycle is mild, and they are constructed of stainless steel. 4. The risk is not significantly changed by replacement of the examinations with visual examinations. This paper will describe in detail the technical arguments under each of these topics, which together form the basis for the code case.

Methods to Define Failure Probability for Power Plant Heat Exchangers

2017 ◽  
Author(s):  
Carolyn J. John ◽  
Consuelo E. Guzman-Leong ◽  
Thomas C. Esselman ◽  
Sam L. Harvey
Keyword(s):  
Heat Exchanger ◽  
Power Plant ◽  
Fresh Water ◽  
Heat Exchangers ◽  
Power Plants ◽  
Operating Experience ◽  
Operating Conditions ◽  
Microbiologically Influenced Corrosion ◽  
Water Steam ◽  
Over Time

In response to the technical challenges faced by aging plant systems and components at nuclear power plants (NPP), the Electric Power Research Institute (EPRI) has a product entitled Integrated Life Cycle Management (ILCM). The ILCM software is a quantitative tool that supports capital asset and component replacement decision-making at NPPs. ILCM is comprised of models that predict the probability of failure (PoF) over time for various high-value components such as steam generators, turbines, generators, etc. The PoF models allow the user to schedule replacements at the optimum time, thereby reducing unplanned equipment shutdowns and costs. This paper describes a mathematical model that was developed for critical heat exchangers in a power plant. The heat exchanger model calculates the probability of the tubes, shell, or internals failing individually, and then accumulates the failures across the heat exchanger sub-components. The dominant degradation mechanisms addressed by the model include stress corrosion cracking, wear, microbiologically influenced corrosion, flow accelerated corrosion, and particle-induced erosion. The heat exchanger model combines physics-based algorithms and operating experience distributions to predict the cumulative PoF over time. The model is applicable to shell and tube heat exchangers and air-to-water heat exchangers. Many different types of fluids including open cycle fresh water, closed cycle fresh water, sea water, brackish water, air, closed cooling water, steam, oil, primary water, and condensate are included. Examples of PoF over time plots are also provided for different fluid types and operating conditions.

Thermal-Hydraulic Character Effect Factors Analysis for Passive Residual Heat Removal System of Sodium-Cooled Fast Reactor

2012 ◽  
Author(s):  
Xiaolong Sun ◽  
Minjun Peng ◽  
Genglei Xia
Keyword(s):  
Heat Exchanger ◽  
Fast Reactor ◽  
Heat Exchangers ◽  
Heat Removal ◽  
Factors Analysis ◽  
Higher Temperature ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Removal System ◽  
Residual Heat

A theoretical investigation was performed on the passive residual heat removal system (PRHRs) of China Experimental Fast Reactor (CEFR). The system contained two air heat exchangers, two independent heat exchangers and several pipes. By means of JTOPMERET code developed by GSE, the thermodynamic model of the system had been built. Based on the model, transient characteristics of the PRHRs have been discussed in detail. Through the calculation analysis, it is then indicated that the capacity of removing residual heat was affected by some factors. The faster speed of air in the air heat exchanger, the higher temperature and the larger heat transfer area of the air heat exchanger are favorable to the PRHRs. At the same time, the calculated parameters variation trends are reasonable according to the Experimental data from CEFR.

scholarly journals Heat Transfer Analysis of Passive Residual Heat Removal Heat Exchanger under Natural Convection Condition in Tank

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Qiming Men ◽  
Xuesheng Wang ◽  
Xiang Zhou ◽  
Xiangyu Meng
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Thermal Stratification ◽  
Heat Removal ◽  
Heat Transfer Analysis ◽  
Average Error ◽  
Water Tank ◽  
Shaped Tube ◽  
Heat Transfer Calculation ◽  
Residual Heat

Aiming at the heat transfer calculation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX), experiments on the heat transfer of C-shaped tube immerged in a water tank were performed. Comparisons of different correlation in literatures with the experimental data were carried out. It can be concluded that the Dittus-Boelter correlation provides a best-estimate fit with the experimental results. The average error is about 0.35%. For the tube outside, the McAdams correlations for both horizontal and vertical regions are best-estimated. The average errors are about 0.55% for horizontal region and about 3.28% for vertical region. The tank mixing characteristics were also investigated in present work. It can be concluded that the tank fluid rose gradually which leads to a thermal stratification phenomenon.

scholarly journals OTEC Maximum Net Power Output Using Carnot Cycle and Application to Simplify Heat Exchanger Selection

Entropy ◽  
2019 ◽  
Vol 21 (12) ◽  
pp. 1143 ◽  
Author(s):  
Kevin Fontaine ◽  
Takeshi Yasunaga ◽  
Yasuyuki Ikegami
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Pressure Drop ◽  
Power Output ◽  
Heat Exchangers ◽  
Power Plants ◽  
Seawater Temperature ◽  
Carnot Cycle ◽  
Ocean Thermal Energy ◽  
Net Power Output

Ocean thermal energy conversion (OTEC) uses the natural thermal gradient in the sea. It has been investigated to make it competitive with conventional power plants, as it has huge potential and can produce energy steadily throughout the year. This has been done mostly by focusing on improving cycle performances or central elements of OTEC, such as heat exchangers. It is difficult to choose a suitable heat exchanger for OTEC with the separate evaluations of the heat transfer coefficient and pressure drop that are usually found in the literature. Accordingly, this paper presents a method to evaluate heat exchangers for OTEC. On the basis of finite-time thermodynamics, the maximum net power output for different heat exchangers using both heat transfer performance and pressure drop was assessed and compared. This method was successfully applied to three heat exchangers. The most suitable heat exchanger was found to lead to a maximum net power output 158% higher than the output of the least suitable heat exchanger. For a difference of 3.7% in the net power output, a difference of 22% in the Reynolds numbers was found. Therefore, those numbers also play a significant role in the choice of heat exchangers as they affect the pumping power required for seawater flowing. A sensitivity analysis showed that seawater temperature does not affect the choice of heat exchangers, even though the net power output was found to decrease by up to 10% with every temperature difference drop of 1 °C.

Experimental Study on Natural Convection Heat Transfer Outside Tube Bundle in Space Under Low Temperature Difference

2020 ◽  
Author(s):  
Junxiu Xu ◽  
Ming Ding ◽  
Changqi Yan ◽  
Guangming Fan
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Low Temperature ◽  
Tube Bundle ◽  
Convection Heat Transfer ◽  
Heat Removal ◽  
Natural Convection Heat Transfer ◽  
Convection Heat ◽  
Reactor Pool ◽  
Residual Heat

Abstract The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

Analysis of Heat Transfer and Flow Characteristics of AP1000 Passive Residual Heat Removal Heat Exchanger

2014 ◽  
Author(s):  
Xu Xie ◽  
Changhua Nie ◽  
Li Zhan ◽  
Hua Zheng ◽  
Pengzhou Li ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Temperature Difference ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Heat Removal ◽  
Large Temperature ◽  
Water Tank ◽  
Secondary Side ◽  
Residual Heat

In this paper, the computational fluid dynamics (CFD) method is applied to the thermal-hydraulic analysis, while the porous media model is used to simplify AP1000 passive residual heat removal heat exchanger tube. The temperature as well as flow distribution in the secondary side of the heat exchanger are obtained, aiming at analysis of natural circulation ability. It can be noted that the fluid in the secondary side of heat exchanger moves driven by the effect of thermal buoyancy, forming the natural cycle, which takes away heat in tube bundle region. The heat transfer in water tank is mainly enhanced by vortex and turbulent flow, caused by the large resistance of tube bundle region as well as large temperature difference. This phenomenon is obvious especially for the recirculation of flow near the tube bundle. The enduring change of flow rate and direction enhance the heat transfer. Besides, the big temperature difference helps to increase the driving effect of natural circulation. Consequently, the heat transfer of the tank is enhanced by above mechanism. The results of this study contribute to the capacity analysis of passive residual heat removal of natural circulation system, providing valuable information for safe operation of AP1000.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Tube Heat Exchanger ◽  
Removal Capacity ◽  
Safety Margins ◽  
Final Design ◽  
Safety Features ◽  
Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant 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 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

Enhanced VVER-1000 Fuel Technology and Performance

2009 ◽  
Author(s):  
H. Shah ◽  
R. Latorre ◽  
G. Raspopin ◽  
J. Sparrow
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Safety Level ◽  
And Performance

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.

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