Deuterium Ingress at the Rolled Joints in CANDU Reactors: Where Does it Comefrom and How Can It be Reduced?

2020 ◽  
Author(s):  
Scott Langille ◽  
Christopher Coleman ◽  
Glenn McRae
Keyword(s):  
Stainless Steel ◽  
Heat Transport ◽  
Heavy Water ◽  
Nuclear Reactors ◽  
Hydrogen Isotopes ◽  
Operating Principle ◽  
Small Scale ◽  
Candu Reactors ◽  
Pressure Tubes

Abstract Simple, small-scale, experiments demonstrate the high deuterium concentrations found in the zirconium pressure tubes at CANDU rolled joints comes from the initial as-received protium in stainless-steel end fittings exchanging with deuterium before being gettered by the zirconium. We propose to reduce the concentration of hydrogen isotopes at the ends of pressure tubes in heavy-water nuclear reactors with yttrium getters placed in the outer regions of the stainless-steel end fittings away from the heat-transport heavy water. Simple, small-scale, experiments demonstrate the operating principle showing that yttrium can getter hydrogen isotopes from the zirconium through the stainless steel.

Experiences in Manufacturing and Qualifying of Stainless Steel AISI 403 Grade Materials for PHWR Reactors of NPCIL

2013 ◽  
Vol 794 ◽  
pp. 194-200
Author(s):  
J.K. Jain ◽  
Sundar Singh ◽  
M.K. Mathur ◽  
N.P. Srivastava ◽  
K.P. Dwivedi
Keyword(s):  
Stainless Steel ◽  
Heavy Water ◽  
Elevated Temperatures ◽  
Coefficient Of Thermal Expansion ◽  
Northern Region ◽  
Nuclear Plants ◽  
Pressure Tubes ◽  
Steel Aisi ◽  
Grade Material ◽  
Stainless Steel Aisi

NPCIL uses stainless steel AISI 403 grade material extensively for very critical nuclear components. A very special application of this material is for manufacturing END FITTING BODY of COOLANT CHANNEL ASSEMBLIES of Pressurized Heavy Water Reactor (PHWR) of Nuclear Power Plants. Presently, two types of PHWR Nuclear Plants viz 220 MWe and 540 MWe are under operation in India. Next higher rating viz 700 MWe PHWR Nuclear Plants are at various stages of construction and due for commissioning in near future. There are number of Pressure Tubes in PHWR Nuclear Plants wherein fuel is loaded and are called coolant channel assemblies. End fitting bodies are assembled on either end of PRESSURE TUBES of coolant channel assembly. The Coolant (Heavy water) flows inside these assemblies and collects the heat produced during nuclear reaction in pressure tubes. This heat is utilized to produce Steam in Steam Generators. After giving up heat in Steam Generator, coolant returns to pressure tube and cycle continues. Material for End Fittings has been selected keeping in mind the mechanical properties at room and at elevated temperatures are compatible with mating component which are manufactured using Zircaloy and SS 410. It is also an essential that coefficient of thermal expansion is also compatible for mating material. Indigenization of the material for manufacturing of End Fittings with short delivery period has always been a challenge for commercial Nuclear Reactors. Inspection of these types of critical nuclear reactor components becomes critical due to the high safety standards and complexity in structures. There are very few manufacturers for this material in India. NPCIL took up a challenge for developing a new supplier in Northern Region of India for developing this material. A leading company in northern region of INDIA came forward and took up the challenge and succeeded for this venture. This paper deals with development of vender to manufacture of AISI 403 grade material by melting, refining ( ESR process), forging, heat treatment, testing & provide the guidelines to tackle the issue faced during this developmental work.

CANDU-Like Features in Emerging Small Modular Reactor Designs

2021 ◽  
Author(s):  
Esam Hussein
Keyword(s):  
Heavy Water ◽  
Operating Experience ◽  
Heat Sinks ◽  
Candu Reactors ◽  
Small Modular Reactor ◽  
Large Heat ◽  
Reactor Shutdown ◽  
To Receive

Abstract Several small modular reactor (SMR) designs are emerging, but only the CANDU Small Modular Reactor and a couple of Indian designs incorporate the familiar features of the larger CANDU-reactors. This paper shows that while the CANDU concept did not seem to receive wider attention among SMR designers, it has influenced a few. The paper discusses how the CANDU operating experience can aid in the construction and operation of some SMRs. For example, the concept of passive reactor shutdown by draining the moderator, which was utilized in the early Pickering A units, is adopted in the Copenhagen Atomics Waste Burner; a molten slat (LiF-ThF$ _4 $) heavy-water moderated reactor. The heavy-water and lithium in this salt produce tritium and can benefit from the CANDU experience in handling tritium. The online refueling of CANDU reactors, their large heat sinks and seamless configuration are also reflected in SMR designs.

scholarly journals Experimental Investigation of Metal-Diamond Thermal Interface Conductance With Different Diamond Surface Terminations

2010 ◽  
Author(s):  
Kimberlee C. Collins ◽  
Gang Chen
Keyword(s):  
Thermal Conductivity ◽  
Heat Transport ◽  
Synthetic Diamond ◽  
Diamond Surface ◽  
Small Scale ◽  
Device Performance ◽  
Thermal Interface ◽  
Single Crystal Diamond ◽  
Heat Spreading ◽  
Resistance To Heat

Synthetic diamond has potential as a heat spreading material due to its uniquely high thermal conductivity. In small-scale devices, interfaces can dominate the resistance to heat transport, and thus play an important role in determining device performance. Here we use transient thermoreflectance techniques to measure the thermal interface conductance at metal-diamond interfaces. We study single crystal diamond samples with various surface terminations. We measure thermal interface conductance values over a range of temperatures from 88 K to 300 K, and find roughly 60 percent higher thermal interface conductance between Al and oxygenated diamond samples as compared to hydrogen terminated samples. The results reported here will be useful for device design and for advancing models of interfacial heat transport.

Experiences in Examining and Dispositioning Flaws in Zr-2.5Nb Pressure Retaining CANDU Reactor Components

2002 ◽  
Author(s):  
Andrew Celovsky ◽  
John Slade
Keyword(s):  
Service Life ◽  
Reactor Core ◽  
Pressure Tube ◽  
Candu Reactors ◽  
Periodic Inspection ◽  
Pressure Tubes ◽  
Surveillance Programs ◽  
Non Destructive ◽  
Limited Service ◽  
Non Destructive Techniques

CANDU reactors use Zr-2.5 Nb alloy pressure tubes, as the primary pressure boundary within the reactor core. These components are subject to periodic inspection and material surveillance programs. Occasionally, the inspection program uncovers a flaw, whereupon the flaw is assessed as to whether it compromises the integrity of the pressure-retaining component. In 1998, such a flaw was observed in one pressure tube of a reactor. Non-destructive techniques and analysis were used to form a basis to disposition the flaw, and the component was fit for a limited service life. This component was eventually removed from service, whereupon the destructive examinations were used to validate the disposition assumptions used. Such a process of validation provides credibility to the disposition process. This paper reviews the original flaw and its subsequent destructive evaluation.

Fatigue Crack Growth Properties of a Cryogenic Structural Steel at Liquid Helium Temperature

1996 ◽  
Vol 118 (1) ◽  
pp. 109-113 ◽  
Author(s):  
Shinji Konosu ◽  
Tomohiro Kishiro ◽  
Ogi Ivano ◽  
Yoshihiko Nunoya ◽  
Hideo Nakajima ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Fatigue Crack ◽  
Austenitic Stainless Steel ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Liquid Helium ◽  
Liquid Helium Temperature ◽  
Small Scale ◽  
Helium Temperature ◽  
Growth Properties

The structural materials of the coils of superconducting magnets utilized in thermonuclear fusion reactors are used at liquid helium (4.2 K) temperatures and are subjected to repeated thermal stresses and electromagnetic forces. A high strength, high toughness austenitic stainless steel (12Cr-12Ni-10Mn-5Mo-0.2N) has recently been developed for large, thick-walled components used in such environments. This material is non-magnetic even when subjected to processing and, because it is a forging material, it is advantageous as a structural material for large components. In the current research, a large forging of 12Cr-12Ni-10Mn-5Mo-0.2N austenitic stainless steel, was fabricated to a thickness of 250 mm, which is typical of section thicknesses encountered in actual equipment. The tensile fatigue crack growth properties of the forging were examined at liquid helium temperature as function of specimen location across the thickness of the forging. There was virtually no evidence of variation in tensile strength or fatigue crack growth properties attributable to different sampling locations in the thickness direction and no effect of thickness due to the forging or solution treatment associated with large forgings was observed. It has been clarified that there are cases in which small scale yielding (SSY) conditions are not fulfilled when stress ratios are large. ΔJ was introduced in order to achieve unified expression inclusive of these regions and, by expressing crack growth rate accordingly, the following formula was obtained at the second stage (middle range). da/dN = CJ ΔJmJ, CJ = AJ/(ΔJ0)mJ, where, AJ = 1.47 × 10−5 mm/cycle, ΔJ0 = 2.42 × 103N/m.

Analysis of Pressure Suppression Pool Swell Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
B. K. Rakshit ◽  
P. Mukhopadhyay
Keyword(s):  
Operating System ◽  
Heat Transport ◽  
Transport System ◽  
Heavy Water ◽  
Computer Software ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Suppression Pool

Abstract The present paper discusses the development of a computer software or code for a best-estimate analysis of Pressure Suppression Pool Hydrodynamics in a Pressurized Heavy Water Reactor (PHWR) system during a Loss-of-Coolant Accident (LOCA) at the primary heat transport system. The software has been developed on Microcomputers, namely, PC-XT or AT (286) under MS-DOS operating system.

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