In this work, a two-dimensional (2D) geometrically-accurate model of the TN-32 cask is generated in ANSYS/Fluent to investigate the effect of backfill gases and their pressures on the peak cladding temperature (PCT). This model is similar to the cask being used in high-burnup (HBU) spent fuel data project lead by the Electric Power Research Institute (EPRI). Helium, nitrogen, argon, and water vapor fill gases are investigated at pressures ranging from atmospheric (∼10 5 Pa) to 100 Pa. Steady-state computational fluid dynamics (CFD) simulations that include the effect of gas rarefaction (temperature-jump) at the gas-solid interfaces are conducted. The PCT as a function of heat generation rate and pressure is reported as well as the heat generation rate that brings the cladding temperature to the radial hydride formation limit. The results show that there are competing effects between the temperature-jump and the thermal conductivity of the gas to increase the fuel rods’ temperature. The low pressures increased the PCT, with the increase being most significant for the helium backfill.

The core of a CANDU (CANada Deuterium Uranium) pressurized heavy water reactor includes several hundred horizontal fuel channels that pass through a calandria vessel containing the heavy water moderator. In each fuel channel, annulus spacers are used to maintain the gap between the cold calandria tube and the hot pressure tube, a pressurized vessel containing the nuclear fuel in contact with heavy water coolant. In order to carry the loads between the pressure tube and calandria tube, the annulus spacers are required to possess adequate structural strength throughout the operating life of the reactor. The Inconel X-750 spacers used in some reactor units are susceptible to irradiation induced degradation. As irradiation fluence increases with operating time, material embrittlement has been observed due to helium bubble formation in the X-750 spacer material. An engineering approach for assessing the structural strength of CANDU annulus spacers has been recently developed. When the ductility of the material is relatively low, the region susceptible to fracture under applied tensile stress may be adequately idealized as a strip-yield process zone surrounded by elastic material and associated with restraining stress. The engineering approach is based on applying the strip-yield process zone methodology to fracture at a nominally smooth surface. Finite element modeling was undertaken to simulate the strip-yield based fracture process zone. The finite element analyses and results are presented in this paper. The finite element results verify the engineering equations developed to assess the structural strength of annulus spacers.

Forged stainless steels are commonly used for the containment of hydrogen isotopes and fracture toughness properties are needed for structural integrity assessments. In this study, the effects of hydrogen and tritium precharging on the fracture-toughness properties of Types 316L and 304L stainless steel forgings were measured. The purpose of the study was to evaluate hydrogen and tritium effects on fracture toughness properties of: (1) Type 316 stainless steel stem-shaped and cup shaped forgings; and (2) Type 304L cylindrical block forgings with two different yield strengths. Arc-shaped fracture toughness specimens were cut from the forgings and precharged by exposing the specimens to hydrogen or tritium gas at 623K and 34.5 MPa. Tritium precharged specimens were aged at 193 K for 45 months prior to testing to build-in helium-3 from tritium decay. In the as-received condition, the J-Integral fracture toughness of the stem, cup, and block forgings were very high and exceeded 1200 kJ/m2 on average. The fracture toughness of specimens cut from the low yield strength Type 304L stainless steel block forging had the highest fracture toughness values and Type 316L stainless steel cup forging had the lowest. The reduced fracture toughness values were attributed to the large strain required to produce the cup forging and its high yield strength. Hydrogen precharging reduced the fracture toughness of the stem, cup, and block forgings to values between 34%–51% of a baseline value which was taken to be the fracture toughness value of the low yield strength block forging. Tritium precharging reduced the fracture-toughness values more than hydrogen precharging because of the effects of helium from radioactive decay of tritium. The fracture-toughness properties of tritium-precharged forgings ranged from 12% to 23% of the baseline values. In general, Type 316L stainless steel was more resistant to toughness reductions by hydrogen or tritium (and decay helium) than Type 304L stainless steel. Yield strength had only minor effects on fracture toughness for the precharged steels.

In the piping fabrication industry Gas Tungsten Arc Welding (GTAW) is used extensively for welding of stainless and high alloy steel pipes, mainly for root pass of single side joints. GTAW process uses an inert gas for both shielding and backing of the weld. Backing gas also known as purge gas, is provided mainly to prevent oxidation of root pass weld from inside the pipe. Typically, Argon or Argon mixture (i.e. Argon with Helium, etc.) are used as backing gas. Argon gas, is colorless, odorless and heavier than air. Many safety incidents have occurred in the Oil and Gas industry (mainly during construction phase) due to asphyxiation. This has typically occurred when a person (usually a welder) goes inside the piping (a confined space) to inspect the root pass weld to ensure it will pass radiography test. Construction industry has failed to fully tackle this issue in a meaningful way and as a result, some significant number of fatalities occur each year. On several occasions, such incidents have resulted in multiple fatalities when coworkers enter the confined space to rescue and become a victim by themselves. There are a multitude of reasons why people put themselves at risk in this way. Reasons given for a reluctance to use alternative techniques that eliminates backing gas are many and varied. Alternate welding techniques without backing gas were developed many years ago. However, they are still not being used extensively for many reasons and apprehensions. Using these alternate welding techniques with right controls in place, can eliminate the hazard associated with backing gas and improve safety and maintain quality. This paper discusses and evaluates the alternate welding techniques (which do not require backing gas), limitations, welding procedure qualification and inspection requirements.

The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainless-steel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.

Radiation-induced degradation of reactor pressure vessels and internals is a concern to the aging nuclear fleet and welding solutions will be required if repair of these irradiated components is deemed necessary. However, the weldability of highly irradiated austenitic materials is significantly diminished due to the presence of irradiation induced helium in the material matrix. Helium-induced weld cracking is a complex phenomenon that is related to the concentration of helium, the heat input from the welding process, and stresses generated during cooling of the weld. During conventional high heat input welding methods such as gas tungsten arc welding, helium bubbles can coalesce and grow on base metal grain boundaries within the heat-affected zone which subsequently causes intergranular cracking. The objective of this work was to obtain weldability data by characterizing welds made on highly activated, neutron irradiated Type 304 stainless steel containing both radiation-induced helium and microstructural damage such as void swelling. All irradiated materials welding was performed inside a Westinghouse hot cell utilizing a pulsed Nd:YAG laser with welds made on three rectangular samples of highly activated Type 304 stainless steel. The rectangular samples were cut and milled in-cell from sections previously obtained from two neutron reflector hex blocks. The hex blocks are U.S. Department of Energy owned material and were irradiated for approximately 13 years in the EBR-II sodium cooled fast reactor from 1982 until 1995. The three samples selected for welding have nominal damage doses of approximately 0.4, 11, and 28 dpa with corresponding estimated helium contents of 0.2, 3 and 8 appm helium, respectively. A number of different weld parameter sets were utilized and included variations of travel speed, wire feed speed and lens-to-work distances. The parameter sets allowed for a range of effective weld heat input levels to be compared. Single pass and multiple pass as well as wire fed and autogenous welds were made. This paper presents the results from post-weld evaluations performed on the three welded irradiated samples, focusing on the reduced tendency for cracks to form adjacent to the weld as a function of weld parameters, lens-to-work distance and helium content.

Previously, in Europe, in gasket tests mainly Nitrogen was used as the test medium. This medium was easy to handle, safe and cheap to buy. With increased sealing requirements on the sealing elements, however, a test medium had to be used, with which smaller leakage quantities can be reliably measured. In laboratory testing, Helium prevailed, with which in the vacuum method using a mass spectrometer these tightness requirements could be easily achieved. Other media, such as hydrogen or Methane, were not used in standard tests because of their hazard potential and detection limitation. Particularly in performance tests, a practice-oriented medium has always been favored in sealing tests in the USA. Methane is the first choice as the smallest hydrocarbon molecule. In ASME B16.20, a new performance test has been introduced, in which the tightness requirement of spiral wound gaskets in a leakage test with Methane should be checked. And also in the valve test in ISO 15848-1, Methane is defined as an alternative test medium for Helium. In this paper, the difference in the leakage test between Nitrogen, Helium and Methane in terms of handling, use in the test laboratory, the measurement of leakage rates and the detection limits between Nitrogen, Helium and Methane will be shown.

The prediction of gas and liquid leak rate through packed stuffing boxes subjected to gas flow is a subject of very few studies in the literature. For better prediction of leakage, the change of porosity with length due to the non-uniform axial stress must be accounted for. There are few theoretical models on the prediction of leak rates in packing rings with capillary models. However, a model that incorporates the change of the capillary area with stress gives a better prediction. In this paper, the first slip flow model is used to predict gas and liquid flow considering the straight capillaries and capillaries having an area dependent on the axial stress in the packing rings. An approach that uses an analytical-computational methodology based on the number and the size of pores obtained experimentally is adopted to predict gas and liquid leak rates in uniform and non-uniform compressed yarned packings. The Navier-Stokes equations associated with slip boundary condition at the wall are used to predict leakage. Experimental tests with helium, argon, nitrogen and air for gazes and water and kerosene for liquids will be used to validate the models. The porosity parameters characterization will be conducted experimentally with helium at a reference gas at different gland stresses and pressures.

The V26 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) for use on the Phase-2 Explosive Destruction Systems. The vessel was fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the Code Case, is nine (9) pounds TNT-equivalent for up to 637 detonations, limited only by fatigue crack growth calculations initiated from a minimum detectable crack depth. The vessel consists of a cylindrical cup, a flat cover or door, and clamps to secure the door. The vessel is sealed with a metal gasket. The body is a deep cylindrical cup machined from a 316 stainless steel forging. The door is also machined from a 316 stainless steel forging. The closure clamps are secured with four 17-4 PH steel threaded rods with 4140 alloy steel threadednuts on one end and hydraulic nuts on the other. A flange with four high-voltage electrical feedthroughs is bolted to the door and sealed with a small metal gasket. These feedthroughs conduct the firing signals for the high-voltage Exploding Bridge-wire detonators. Small blast plates on the inside of the door protect fluidic components and electrical feedthroughs. A large blast plate provides additional protection. Both vessel door and feedthrough flange employ O-ring seals outside the metal seals in order to provide a mechanism for helium leak checks of the volume just outside the metal seal surface before and after detonation. In previous papers (References 2 and 3), the authors describe results from testing of the vessel body and ends under qualification loads, determining the effective TNT equivalency of Composition C4 (EDS Containment Vessel TNT Equivalence Testing) and analyzing the effects of distributed explosive charges versus unitary charges (EDS Containment Vessel Explosive Test and Analysis). In addition to measurements made on the vessel body and ends as reported previously, bulk motion and deformation of the door and clamping system was made. Strain gauges were positioned at various locations on the inner and outer surface of the clamping system and on the vessel door surface. Digital Image Correlation was employed during both hydrostatic testing and dynamic testing under full-load explosive detonation to determine bulk and bending motion of the door relative to the vessel body and clamping system. Some limited hydrocode and finite element code analysis was performed on the clamping system for comparison. The purpose of this analysis was to determine the likelihood of a change in the static sealing efficacy of the metal clamping system and to evaluate the possibility of dynamic burping of vessel contents during detonation. Those results will be reported in this paper.

During vacuum drying of used nuclear fuel (UNF) canisters, helium pressure is reduced to as low as 67 Pa to promote evaporation and removal of remaining water after draining process. At such low pressure, and considering the dimensions of the system, helium is mildly rarefied, which induces a thermal-resistance temperature-jump at gas–solid interfaces that contributes to the increase of cladding temperature. It is important to maintain the temperature of the cladding below roughly 400 °C to avoid radial hydride formation, which may cause cladding embrittlement during transportation and long-term storage. Direct Simulation Monte Carlo (DSMC) method is an accurate method to predict heat transfer and temperature under rarefied condition. However, it is not convenient for complex geometry like a UNF canister. Computational Fluid Dynamics (CFD) simulations are more convenient to apply but their accuracy for rarefied condition are not well established. This work seeks to validate the use of CFD simulations to model heat transfer through rarefied gas in simple two-dimensional geometry by comparing the results to the more accurate DSMC method. The geometry consists of a circular fuel rod centered inside a square cross-section enclosure filled with rarefied helium. The validated CFD model will be used later to accurately estimate the temperature of an UNF canister subjected to vacuum drying condition.

In this work, a geometrically-accurate two-dimensional (2D) computational fluid dynamic (CFD) model of a used nuclear fuel cask, that can contain up to 32 pressurized water reactor (PWR) used nuclear fuel (UNF) assemblies, is constructed. This model is similar to the TN-32 cask employed in the ongoing high-burnup (HBU) Spent Fuel Data Project lead by the Electric Power Research Institute (EPRI). This model is used to predict the peak cladding temperature under vacuum drying conditions. Due to the symmetry of the cask, only one-eighth of the cross-section is modeled. Steady-state simulations that include the temperature-jump boundary conditions at the gas-solid interfaces are performed for different heat generation rates in the fuel regions and a range of dry helium pressures, from ∼10 5 to 100 Pa. These simulations include conduction within solid-gas regions and surface-to-surface radiation across all gas regions. The peak cladding temperatures are reported for various heat generation rates and rarefaction conditions, along with the maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit. The results showed that the decrease of helium pressure significantly increased the temperature of the cladding material compared to the atmospheric pressure condition.

Vacuum drying of nuclear fuel canisters may cause the temperature of fuel assemblies to considerably increase due to the effect of gas rarefaction at low pressures. This effect may induce a temperature-jump at the gas-solid interfaces. It is important to predict the temperature-jump at these interfaces to accurately estimate the maximum temperature of the fuel assemblies during vacuum drying. The objective of this work is to setup a concentric cylinders experimental apparatus that can acquire data to benchmark rarefied gas heat transfer simulations, and determine the temperature-jump coefficient at the interface between stainless steel surface and helium gas. The temperature-jump is determined by measuring the temperature difference and heat flux across a 2-mm gap between the concentric cylinders that contains rarefied helium and compare the results to analytical calculations in the slip rarefaction regime.

The hydrogen environment embrittlement (HEE) on SUS 630 was evaluated by enclosing the hydrogen gas into a hole of the specimen at 77 K, 120 K, 190 K, 230 K, 260 K, 300 K, and 373 K. The pressure range of hydrogen gas was 10 MPa to 70 MPa. The slow strain-rate tensile (SSRT) tests were carried out and the relative reduction of area (RRA) was calculated by the ratio of the reduction of area (RA) in hydrogen gas and the RA in helium gas. The RA and tensile strength were smaller in the higher pressure of hydrogen gas. The RRA between 190 K and 260 K was almost zero, however, the HEE reduces at 77 K and 373 K. There was almost no HEE after changing to helium gas at the stress of just below the fractured stress in hydrogen during the SSRT tests.

This paper presents the design and analysis of the Hydrogen Transfer Lines (HTL) in the Spallation Neutron Source Facility (SNS), Oak Ridge National Laboratory in Oak Ridge, Tennessee. SNS is a Neutron Accelerator providing intense pulsed neutron beams for scientific research into the behavior of materials. The use of supercritical Hydrogen provides moderation of these neutron beam lines. This is facilitated by the HTL, which consists of six coaxial tubes transporting supercritical hydrogen to and from the Moderators surrounding the Mercury Target inside the Core Vessel (CV). Each Hydrogen line is surrounded by a vacuum tube in order to preserve the supercritical Hydrogen temperature of −425°F (−254° C/19 K). An inert Helium tube and a Water Jacket surround the Outer Vacuum Jacket. The Hydrogen boundary is considered the primary containment boundary, with the vacuum tubing being the secondary containment boundary both considered to be safety significant. The HTL assembly tubing must be qualified as Performance Category 3 to the requirements of Department of Energy Standard 1020-1994 for Natural Phenomena Hazards. This requires a detailed dynamic analysis. The HTL must also be qualified to ASCE 43-05 for Limit State consideration for abnormal events and ASME B31.3 for sustained and occasional (upset) loadings. The design basis includes consideration of normal sustained loadings such as deadweight and pressure, as well as thermal expansion/contraction, a Design Basis Earthquake (DBE) with associated Anchor Movements (SAM) and the potential physical interaction between coaxial tubes. Intermittent spacers are provided between some of the tubes, providing a functional interspace between the Hydrogen tubing and the Vacuum Jackets. The analytical methodology involves supplementing unaddressed criteria per ASME B31.3 thus requiring the application of more detailed rules found in ASME Section III, NC-3600, ASCE/SEI 43 – 05 as well as Regulatory Requirements such as USNRC Regulatory Guide 1.61 and 1.92. To assess the potential of a postulated failure of the primary Hydrogen boundary inside the Core Vessel, NUREG 0800, Section 3.6.3, Leak-Before-Break Evaluation Procedures are employed.

Archival materials test data on austenitic stainless steels for service in high pressure hydrogen gas has been reviewed. The bulk of the data were from tests conducted prior to 1983 at the Savannah River Laboratory, the predecessor to the Savannah River National Laboratory, for pressures up to 69 MPa (10,000 psi) and at temperatures within the range from 78 to 400 K (−195 to 127 °C). The data showed several prominent effects and correlations with test conditions: • There was a significant reduction in tensile ductility as measured by reduction of area or by the total elongation with hydrogen. Hydrogen effects were observed when the specimens were tested in the hydrogen environment, or the specimens were precharged in high pressure hydrogen and tested in air or helium. • There was a significant reduction in fracture toughness with hydrogen (and sometimes in tearing modulus which is proportional to the slope of the crack resistance curve). • The effects of hydrogen on ductility can be correlated to the nickel content of the iron-chromium-nickel steels. The optimal nickel content to retain the high tensile ductility in these alloys was 10 to at least 20 wt. %. • The effects of hydrogen can be correlated to the grain size. Large grain sizes exhibited a greater loss of ductility compared to small grain sizes. The Savannah River Laboratory test data, especially those not readily available in the open literature, along with the sources of the data, are documented in this paper.

During vacuum drying of used nuclear fuel canister, helium pressure is decreased to as low as 67 Pa to promote evaporation and removal of water remaining in the canister following draining operation. At low pressures associated with vacuum drying, there is a temperature jump (thermal resistance) between the solid surfaces and helium in contact with them. This temperature jump increases as the pressure decreases (rarefied condition), which contributes to the fuel assembly’s temperature increase. It is important to keep the temperature of the fuel assemblies below 400°C during vacuum drying to ensure their safety for transport and storage. In this work, an experimental apparatus consisting of a 7×7 array of electrically heated rods maintained between two spacer plates and enclosed inside a square cross-section stainless steel pressure vessel is constructed to evaluate the temperature of the heater rods at different pressures. This geometry is relevant to a BWR fuel assembly between two consecutive spacer plates. Thermocouples are installed in each of the 49 heater rods, spacer plates and enclosure walls. They provide a complete temperature profile of the experiment. Different pressures and heat generation relevant to vacuum drying conditions are tested. The results showed that the maximum temperature of the heater rods increases as the pressure decreases. The results from these experiments will be compared to computational fluid dynamics simulations in a separate work.

Computational fluid dynamics simulations of a 7×7 array of heated rods within a square-cross-section enclosure filled with rarefied helium are performed for heat generation rates of 50 W and 100 W and various helium pressures ranging from 10 5 to 50 Pa. The model represents a section of nuclear fuel assembly between two consecutive spacer plates inside a nuclear canister subjected to during vacuum drying process. A temperature jump model is applied at the solid-gas interface to incorporate the effects of gas rarefaction at low pressures. The temperature predictions from simulations are compared to measured temperatures. The results showed that when helium pressure decreased from 10 5 to 50 Pa, the maximum temperature of the heater rod array increased by about 14 °C. The temperatures of the hottest rod predicted by simulations are within 4°C of the measured values for all pressures. The random difference of simulated rod temperatures from the measured rod temperatures are 3.33 °C and 2.62 °C for 100 W and 50 W heat generation rate.

A primary repair option of Light Water Reactors (LWR) components is welding. However, it is known that welding on steels that have been exposed to neutron irradiation [i.e. Reactor Pressure Vessels (RPV) in PWR and Reactor Internals in BWR] can result in Helium Induced Cracking (HeIC). Helium forms from neutron transmutation reactions of Boron (B) and Nickel (Ni) during operation of the plant. In order to address this issue and establish verified methods for weld repair of irradiated RPV and Reactor Internals materials in Japanese power plants, an investigation denominated WIM ( Welding of Irradiated Materials ) Project was conducted; the WIM project was carried out between the years of 1997 and 2004 in an analytical conservative manner, correlating the results of weld repaired irradiated materials with the concentration of helium and the heat input used while welding. It was concluded that, under determined conditions, the irradiated materials were able to be successfully welded in accordance with the requirements established in both the JSME and ASME Code Cases. In the light of such discovery, the necessity of establishing and a new code case and revising the standard JSME Rules on Fitness-for-Service concerning the weld repair of irradiated RPV and Reactor Internals steels is currently under investigation.

Substantial research has been performed in recent years to determine the effects and feasibility of welding on highly irradiated austenitic materials. This research has been driven by the need to preemptively develop welding techniques capable of repairing highly irradiated light water reactor (LWR) components susceptible to detrimental corrosion and cracking. However, the materials used to fabricate internal LWR components become increasingly difficult to weld with in-service age due to irradiation-induced generation of helium in the material matrix over time. This paper introduces a patent-pending technology that proactively manages the stresses during laser repair welding of highly irradiated reactor internals to avoid the occurrence of intergranular helium-induced cracking. The technology development relied on numerical simulations that made it possible to refine and optimize the innovative welding concept and to identify specific process conditions achieving significant reduction of tensile stress (or even formation of compressive stress) near the weld pool in the heat-affected zone on cooling. The candidate welding process conditions identified by the numerical simulations were experimentally tested on stainless steel plates (Type 304L) with a laser welding system purposely designed and engineered to incorporate the proactive stress management concept. In-situ temperature and strain measurement technique based on digital image correlation were applied to validate the numerical simulations.

As the existing light water reactor (LWR) fleet ages, the weldability of structural materials used to construct the reactor pressure vessels (RPVs) and reactor internals is diminished. The decrease in the weldability in austenitic and ferritic materials is attributed to the formation of helium in the material microstructure. Helium (He) generation occurs during the service life of irradiated reactor internals from neutron transmutation reactions of boron and nickel in these materials. Welding on irradiated materials, if performed without appropriate consideration of fluence exposure and helium generation, can result in a heat affected zone cracking phenomenon termed helium induced cracking (HeIC). The heat input associated with welding is a major factor affecting the coalescence of the generated helium along grain boundaries. As the material cools, the tensile stresses generated from welding can cause cracking to occur along grain boundaries weakened by helium bubble coalescence. In some cases, the preferred or only method of repair or replacement of a reactor internal component is welding. For components located in regions of low thermal fluence, the welding process implementation may be relatively straightforward and only heat input control may be required. However, in high thermal fluence regions, weld repair of irradiated reactor internal components is complicated by the presence of high concentrations of helium and significant care must be taken in welding process selection and heat input control. This paper highlights envisioned applications for weld repair on irradiated reactor internals. It also summarizes recently completed guidance published by the EPRI Materials Reliability Program (MRP) and EPRI BWR Vessel and Internals Project (BWRVIP) which provides an improved basis for plants to assess the weldability of components at various locations within the reactor.