bluuebunny/crossref_metadata_2025_split huggingface dataset:
CFD Analyses of the TN-24P PWR Spent Fuel Storage Cask
Authors: Brewster Robert A. , Baglietto Emilio , Volpenhein Eric , Bajwa Christopher S. 
Published: August 2013

Dry storage casks are used to store spent nuclear fuel after removal from the reactor spent fuel pool. Even prior to the Fukushima earthquake of March 2011, dry storage of spent fuel was receiving increased attention as many reactor spent fuel pools near their capacity.
Many different types of cask designs are used, and one representative design is the TN-24P spent fuel cask, a non-ventilated steel cask with a shielded exterior shell and lid. The cask is typically filled with an inert gas such as helium, argon or nitrogen. In this paper, Computational Fluid Dynamics (CFD) calculation results for the thermal performance of the TN-24P cask using the commercial CFD software STAR-CCM+ are presented. Initial calculations employ a common approach of treating the fuel assemblies as conducting porous media with calibrated volume-averaged properties, and comparison to existing measured temperature data shows good agreement.
One of the fuel assemblies is then replaced with a more accurate representation that includes the full geometric detail of the fuel rods, guide tubes, spacer grids and end fittings (flow nozzles), and the results shown are consistent with the initial analysis, but without the need for the assumptions inherent in the porous media approach. This hybrid modeling approach also permits the direct determination of important results, such as the precise location of peak fuel cladding temperatures (PCTs), which is not possible using the more traditional porous media approach.
Numerical Analysis of Bolted Trunnion Systems of Packages for Radioactive Materials
Authors: Sterthaus Jens , Ballheimer Viktor , Kuschke Christian , Wille Frank 
Published: August 2013

Packages for the transport of radioactive material are generally equipped with particular components for crane operations and supporting the package during transport. As an example of such components the bolted trunnion system of heavy packages are analysed in this paper.
According to the German guideline BAM-GGR 012 created by the BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems, the finite element (FE) method is to be used preferably in the analysis of such kind of structures to obtain more accurate and detailed information about their stressing. The FE model of a trunnion system should include the trunnion itself, the bolts and an appropriate part of the packaging wall with necessary contact conditions on all interfaces between these components. The use of solid finite elements, which is generally recommended in the BAM-GGR 012 guideline, leads to local stress and strain fields as a result of the calculation. The modelling strategy, e.g. the way of modelling interaction between the components of the trunnion system, has an essential impact on these local values.
The effect of different assumptions in a FE model is demonstrated in this paper using the calculation results for a reference trunnion system. The approaches for preparation and interpretation of results in connection with local or nominal assessment criteria are discussed as well.
Code Aspects Related to Combination of Stresses due to Hydrogen Detonations: Local Effects
Authors: Minichiello John C. , Ahnert Stephen D. , Ligon Thomas C. , Gross David J. 
Published: August 2013

This paper addresses the local effects of hydrogen detonations inside piping. It is the first in a two-part series of papers which assess the effects of detonations in piping systems relative to ASME Code allowables.
The effects of internal detonations in piping systems are typically separated into two regimes: local effects and system effects. Local effects are often simplistically represented as pure hoop stresses resulting from the pressure acting radially on the inside circumference of the pipe. In reality, the interaction of the pipe wall and the propagating detonation wave is relatively complex, resulting in “waves” or “ripples” in the pipe wall. These areas of local, through-wall curvature lead to substantial axial stresses which may even exceed the hoop stresses. Furthermore, in the elastic regime, there is very little damping present in the pipe wall, leading to numerous stress cycles as the local bending waves move axially along the pipe wall.
Fatigue effects of the combined hoop and axial cycling were evaluated using ASME Code Section VIII, Division 2 fatigue evaluation methodology. Analysis of strain gage data from a number of hydrogen detonation experiments in 2-inch and 4-inch Schedule 40 piping showed that the fatigue damage is generally driven by fewer than 10 large-magnitude fatigue cycles, which account for an average of 75% of the total fatigue damage. However, the results also demonstrate that for two detonation events with similar measured peak hoop or axial strain, the number of fatigue allowable events may vary dramatically depending on the shape of the strain response.
Simplified Stress Linearization Method, Maintaining Accuracy
Authors: Strzelczyk Andrzej , Stojakovic Mike 
Published: August 2013

ASME PVP Code stress linearization is needed for assessment of primary and primary-plus-secondary stresses. The linearization process is not precisely defined by the Code; as a result, it may be interpreted differently by analysts. The most comprehensive research on stress linearization is documented in the work of Hechmer and Hollinger [1]. Recent non-mandatory recommendations on stress linearization are provided in Annex 5A of Section VIII, Division 2 of ASME PVP Code [2]. In the work of Kalnins [3] some linearization questions are discussed in two examples; the first is a plane-strain problem and the second is an axisymmetric analysis of primary plus secondary stress at a cylindrical-shell/flat-head juncture. Paper [3] concludes that for the second example the linearized stresses produced by Abaqus [5] diverge, therefore they should not be used for stress evaluation for this specific case.
This paper revisits the axisymmetric analysis discussed in [3] and attempts to show that the linearization difficulties can be avoided. The paper explains in details the reason for the divergence; the Abaqus program does not linearize all stress components in axisymmetric elements; two stress components are calculated from assumed formulas and all others are linearized. It is shown that when the axisymmetric structure from [3] is modeled with 3D elements, the linearization results are convergent. Further, it is demonstrated that both axisymmetric and 3D modeling produce the same and correct stress Tresca stress, if the stress is evaluated from all stress components linearized, without any further modification.
The stress evaluation of the axisymmetric model of [3] is the primary-plus-secondary-stresses evaluation for which the limit analysis described in [4] cannot be used. The paper shows how the original primary-plus-secondary-stresses problem can be converted into the equivalent primary-stress problem, that is for a problem for which limit analysis can be used; it is further shown how the limit analysis had been used for verification of the linearization results.
Creep Behaviour of P92 and P92 Welds at 675°C
Authors: Tanner David W. J. , Saber Mohammed , Sun Wei , Hyde Thomas H. 
Published: August 2013

The results of an accelerated creep test programme on the leading new build high temperature power plant pipework ferritic steel P92 and two P92 welds are presented.
Tests were performed at 675°C, which is above the service operating temperature range recommended for P92, but allowed for more realistic operating stresses to be used. Comparison with similar tests of P92 at lower temperatures has shown that testing at 675°C produces the same general creep behavior, and can therefore be used for component life assessments in the service operating temperature range.
Axially loaded parent P92 material uniaxial round bar and notched bar specimens, given a heat treatment equivalent to that given post-welding, are compared with weld metal specimens and cross-weld specimens extracted from welds made using both a similar P92 consumable and a dissimilar IN625 nickel alloy consumable. Both welds exhibited typical premature ferritic steel weld failure at the heat-affected zone (HAZ) and parent material interface, known as Type IV.
The creep crack growth behavior of parent material, weld metal and the HAZ was studied using compact tension specimens. Impression testing was performed to determine the relative creep behavior of the HAZ. The HAZ was found to have the highest minimum creep strain rates and creep crack growth rates, indicating the relative weakness of this region.
Submerged Pressure Vessel Testing Hazards
Authors: Montoya John , Ketchum Donald , Edel Matthew 
Published: August 2013

It is common practice to proof test high pressure vessels prior to their use in the field. One technique for leak testing these vessels is submersion in water. A test failure at high pneumatic pressure and can pose several hazards to nearby personnel, such as projectile launch and blast loads. Submerged underwater testing can provide some level of protection from these hazards. However, it is largely unknown how much water cover is needed to prevent a projectile from escaping.
The purpose of this test program was to record the mitigating effects of water on hazards caused by a sudden pressure vessel failure. The test program entails submerging a pressure vessel underwater inside a tank. The vessel is then pressurized to failure, releasing a blast wave and launching a projectile. The event is recorded using high speed photography which is used to observe the effects of the gas release and the projectile motion. A discussion of the test events and associated physics is provided.
Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle
Authors: Zhang Tao , Brust Frederick W. , Wilkowski Gery , Xu Heqin , Betervide Alfredo A. , Mazzantini Oscar 
Published: August 2013

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. The original plant was designed by KWU in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Many times welding residual stresses in nuclear power plant can lead to cracking concerns. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe and nozzle joints with welds.
In this paper, a large diameter Reactor Pressure Vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main weld material is 20MnMoNi55 and ID welded cladding is made of 309L. The special materials and weld geometry will lead to interesting welding residual stress field. In addition, post-weld heat treatment (PWHT) region and boundary conditions could also play an important role in determining welding residual stress fields. Sensitivity analyses were conducted and the technical observation and comments will conclude the paper.
LBB Under Beyond Design Basis Seismic Loading
Authors: Zhang Tao , Brust Frederick W. , Wilkowski Gery , Xu Heqin , Betervide Alfredo A. , Mazzantini Oscar 
Published: August 2013

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference.
This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.
Fitness for Service of Degraded Grade 91 Pipe
Authors: Cohn Marvin J. , Paterson Steve R. 
Published: August 2013

The use of creep strength enhanced ferritic alloys such as Grade 91 in fossil power plants has become popular for high temperature piping applications. Since Grade 91 has higher stress allowables than Grade 22, a designer can specify thinner component wall thicknesses, resulting in lower through-wall thermal stresses during transient events and lower material and piping support costs.
During the past two decades, Grade 91 has been used successfully in fossil power plants. However, this alloy has had some incidents of non-optimal weldment microstructure. In this case study, Brinell hardness tests of an ASME A182 Grade F91 (F91) wye block, including upstream and downstream F91 spools, revealed several readings of soft material, as low as 168HB. A study of creep rupture tests of degraded Grade 91 specimens revealed that the lower bound creep rupture curve of the degraded Grade 91 material is above the average creep rupture curve of Grade 22 material for the range of the specific piping operating stresses.
Based on the empirical evidence that the average Grade 22 material creep rupture curve is conservative for the creep rupture of degraded Grade 91 material, a life consumption evaluation was performed for the degraded Grade 91 weldments using Grade 22 creep rupture properties. A life fraction analysis was performed considering the redistributed maximum principal stresses, based on simulation of piping displacements obtained from the hot and cold walkdowns. This study also considered the recent history of the specific piping system operating pressures and temperatures.
This study also considered dissimilar metal welds, from ASME A182 Grade F91 (F91) to ASME A335 Grade P22 (P22) materials. It was determined that the Grades F91-to-F91 weldments had about 30% life consumption and the remaining lives were at least 7 years. The Grades F91-to-P22 weldments had less than 40% life consumption and the remaining lives were at least 15 years.
Comparison of ASME B31.1 Sustained Load Stresses to Corresponding Tresca Stresses
Authors: Cohn Marvin J. 
Published: August 2013

Conventional United States designs of high energy piping (HEP) systems use the American Society of Mechanical Engineers (ASME) B31.1 Power Piping Code. The analytical methodology in this code is based on linear elastic beam theory. The ASME B31.1-2010 Power Piping Code (Code) [1] recommends Equation 15 to calculate the piping stress due to sustained loads. Many practitioners believe that the sustained load stress (SL) results using Equation 15 are not significantly less than using a Tresca methodology for the same set of forces and moments. This paper provides a comparison of the ASME B31.1 SL stresses to the corresponding Tresca stresses in parent material, based on empirical HEP system stress analyses. The results of three piping system evaluations are considered, including examples of longitudinal stress lower than the circumferential stress and examples where the longitudinal stress is greater than the circumferential stress.
This study considers the elastic primary stresses on the outside surface of the pipe, prior to any creep redistribution. At locations where the longitudinal stress is greater than the circumferential stress, the SL stress is nearly the same as the elastic Tresca stress. At locations where the longitudinal stress is considerably less than the circumferential stress, the SL stress is considerably less than the elastic Tresca stress. This conclusion is due to the fact that the SL stress is primarily governed by longitudinal loading.
The paper also considers inelastic primary stresses, after complete creep redistribution. For piping materials operating in the creep regime, the axial and circumferential pressure stresses are eventually redistributed and are maximum at the outer surface of the pipe. After several years of operation, the Code SL stresses and elastic Tresca stresses are significantly less than the inelastic Tresca stresses. Consequently, the use of SL stresses and elastic Tresca stresses for estimating component inelastic primary stresses would be nonconservative.
Graded Approach to Small Modular Reactor Implementation
Authors: Culp William G. , Cintron-De Leon Brendaly 
Published: August 2013

The regulatory process covers health and safety issues in the context of large power reactors; this paper addresses how requirements on designers and owners might be scaled for small modular reactors (SMR) in proportion to the risks for the type of technology.
Light water SMRs are designed to incorporate many of the fuel and passive safety features already described in the Design Control Documents and Final Safety Analysis Report (FSARs) for large power reactors. Regulations are written to principally address large and advanced power reactors. This paper proposes a several step process to simplify the process of getting design certification, a license to manufacture LWR type SMR, construct a SMR at multiple sites, and accelerate an applicant’s Early Site Review and Combined Operating License process.
This paper looks at issues identified by the NRC Commissioners in SECY 10, 0034, 3/28/10, Potential Policy, Licensing, and Key Technical Issues for Small Modular Reactor Designs, from the industry perspective. Since the SMR issues are less complex and involve less risk than the corresponding issues for large reactors, this paper describes theoretical means of reducing the scope, schedule, and cost of a SMR.
Advantages of Bladder Surge Tanks in Pipelines
Authors: Young Winston B. 
Published: August 2013

The paper advocates the use of bladder surge tanks to control pressure surges in pipelines created by transient operations. Transient pressures are created as a result of pumps starting and stopping as well as the opening and closing of valves. This paper describes why the utilization of bladder surge tanks is the best solution to transient operations. A bladder surge tank allows piping systems to meet pressure piping codes such as ASME B31.3 and B31.4. Controlling pressure transients reduces pressure stresses, thereby extending the pipeline’s operating life. This paper compares the use of a bladder surge tank to other methods to attenuate pressure transient in pipelines.
Generally, water and hydrocarbon pipelines operate safely throughout the world, but failure can occur as a result of pressure transients created by certain operating conditions. In addition, pipe materials are weakened as a result of years of operation and exposure to corrosion, pressure cycling fatigue and over-pressurization.
The usable life of piping materials can be significantly reduced by cyclic loading and surge pressure overstressing the material. Therefore, controlling pressure surges will increase pipeline life and reliability. This can be achieved by installing surge control equipment at the location of the transient event.
Probabilistic Parameter Sensitivity Study on Design Fatigue Curve for Carbon and Low-Alloy Steels
Authors: Takanashi Masahiro , Maeda Jyunki , Sakai Shinsuke 
Published: August 2013

In a previous paper[1], we discussed the margins of the design fatigue curve in ASME Boiler and Pressure Vessel Codes Section III using a probabilistic approach. Limit state functions of the design fatigue curve for carbon and low-alloy steels were proposed in order to investigate the physical implication of the design margin. In the limit state functions, four parameters, namely, fatigue data scatter, size effect, surface roughness, and applied stress were taken into consideration as random variables. Based on the limit state functions, reliability index contours of the design fatigue curve were obtained together with the partial safety factors (PSFs).
Among these partial safety factors, the fatigue strength (or life) and the applied stress were predominant. The other parameters, the PSFs for the size effect and the surface roughness were small enough and almost constant. The parameter sensitivity, however, remains unknown and whether these parameters should be treated as constants or as random variables is also not clear.
In this study, a probabilistic parameter sensitivity study on the design fatigue curve was conducted to discuss how the design fatigue curve should be. Two sensitivities were computed. One is the rate sensitivity indicating the contribution of the mean value of the parameter in the probability of failure. The other is the probabilistic sensitivity to parameter dispersion. The parameters for the size effect and the surface roughness showed lower probabilistic sensitivities. This result suggests that the parameters can be considered as constants. In general, a higher number of parameters in a probabilistic model leads to more uncertainty and the design concept tends to be more conservative. We, therefore, proposed to deal with the parameters of the size effect and the surface roughness separately in the design fatigue curve to eliminate conservativeness.
Simulation of Notch Tip Stress/Strain Response of Zirconium Pressure Tube Material
Authors: Reinhardt Wolf , Doddihal Preeti , Ranganath Sampath 
Published: August 2013

The Fitness-for-Service assessment of Zr 2.5Nb pressure tubes in CANDU reactors requires the evaluation of fatigue. As-designed, the pressure tubes do not contain notches, but in operation various degradation mechanisms, such as fretting and crevice corrosion, can generate flaws with small root radii. Since these flaws are exposed to the coolant, an assessment of environmental effects on fatigue life is needed, and fatigue tests are being conducted to obtain the fatigue curve.
The pressure tubes are located in areas of high neutron flux, and thus the material is subject to irradiation effects. Since tests of irradiated material in reactor coolant environment are very difficult to conduct, a mechanistic understanding of fatigue at notches in Zr 2.5Nb material is sought, to allow the best possible use of fatigue tests of irradiated material in air and unirradiated material in water.
As part of the effort to develop a mechanistic understanding, continuum-level simulations of the mechanical behaviour of unirradiated and irradiated Zr 2.5Nb material were performed using finite element analyses. Both monotonic and cyclic runs were conducted to investigate the difference between first-cycle and shakedown behaviour near the notch tip. To increase confidence in the result, finite element notch strain predictions were benchmarked with a Neuber notch strain model. The paper discusses observations and possible consequences of these simulations.