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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

PCI analysis of a commercial PWR using BISON fuel performance code

Through the early 2000’s, nuclear fuel cladding failure resulting from pellet-cladding interaction in commercial pressurized water reactor fuel was managed successfully by incorporating power ramp restrictions. Within the past decade, however, cladding failures have again begun to appear, often correlated to missing pellet surface defects, and further aggravated by modern reactor operating strategies demanding higher fuel duty operations. In response, renewed efforts in fuel behavior modeling and simulation are being pursued to improve our understanding of the conditions that can lead to missing pellet surface related cladding failure. This paper describes a methodology used to model pellet-cladding interaction related failures observed in commercial pressurized water reactors using the BISON fuel performance code developed at Idaho National Laboratory. In order to accomplish this goal, a three-step process is followed: 1) A failure-stress threshold is developed by using BISON code to model ramp test with known pellet-clad interaction related failures, 2) BISON analysis is applied to a commercial pressurized water reactor cycle startup with one rod confirmed to contain a missing pellet surface defect, and 3) an missing pellet surface sensitivity study of a cycle startup of the same commercial pressurized water reactor which contained a failed fuel rod of un-identified type. The analysis results described in this paper serve to illustrate the BISON code capabilities for modeling pressurized water reactor fuel subjected to higher duty power events.

Nuclear Science & Technology↗
DOE OSTI2018

Methodology for Stress-Controlled Fatigue Test Under In-Air and Pressurized Water Reactor Coolant Water Condition and to Evaluate the Effect of Pressurized Water Reactor Water and Loading Rate on Ratcheting [Methodology for Stress-Controlled Fatigue Test Under In-Air and PWR Coolant Water Condition and to Evaluate the Effect of PWR Water and Loading Rate on Ratcheting]

Here, this work investigates the behavior of 316 stainless steel (SS) under stress-controlled low cycle fatigue loading. Several fatigue experiments are conducted under different environment such as in air at 300°C and primary loop water conditions for a pressurized water reactor (PWR). Two different loading conditions are also employed to examine the effect of stress rate on material hardening and ratcheting. During PWR water test, actuator position measurements are used to determine the strain of the specimen. Under PWR environment, 316 SS is found to ratchet to a significantly greater degree compared with in air. At slow stress rate, higher amount of cyclic hardening is observed in 316 SS, and slow stress rate increases the rate of ratcheting. Results also indicate that 316 SS exhibits asymptotic strain response at higher stress loading which can cause material to behave very differently under same stress cyclic loading.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Interface tracking simulations of bubbly flows in PWR relevant geometries

[en] Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D_h) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in the smaller case and 262 bubbles in the larger case. In the larger simulation case the size of the resolved bubbles is 0.65 mm in diameter, and the bulk mesh cell size is about 30 microns. Those large-scale simulations provide new level of details previously unavailable and were enabled by the excellent scaling performance of our two-phase flow solver and access to the state-of-the-art supercomputing resources. The presented simulations used up to 256 thousand processing threads on the IBM BG/Q supercomputer “Mira” (Argonne National Laboratory).

Nuclear Science & Technology↗
DOE OSTI2017

Stress corrosion crack initiation of alloy 600 in PWR primary water

Stress corrosion crack (SCC) initiation of three mill-annealed (MA) alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular (IG) attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and IGSCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in MA alloy 600 are discussed. IN PRESS, CORRECTED PROOF, 05/02/2017 - mfl

alloy↗
DOE OSTI2019

Probability of Pinching Failure in PWR SNF after Long-Term Storage

A probabilistic assessment was performed on pinching failure of PWR SNF rod cladding under accidental SNF cask transport conditions. a set of 3,000 rod pinching scenarios was analyzed, with different geometrical, material, and loading parameters, using a Latin Hypercube Sampling (LHS). The investigated accidents include train derailments leading to cask collisions or cask drops. Pinching failure was related to cladding offset strain, defined as the plastic deformation interval in radial cladding direction normalized with respect to the outer cladding diameter. Finite element (FE) models were used to compute SNF rod deformations and the offset strain demand OSD in each of the assessed pinching scenarios. The numerical models account for the contribution of the degraded fuel pellets to rod pinching stiffness. The offset strain capacity OSC was computed using regression and interpolation models based on ring compression (RCT) test results, and account for hydride-related effects on cladding ductility. The results show a low probability of pinching failure. Hydride-related cladding embrittlement affects the structural integrity of PWR SNF rods after long-term dry storage of up to 300 years. Radial hydride patterns or large amounts of circumferentially oriented hydrides reduce the cladding ductile deformation capability when pinched at low material temperatures. The probability of hydride-related pinching failure of PWR cladding is estimated after long-term dry storage to address: How do PWR SNF rod conditions like hydrogen content, vacuum drying peak cladding hoop stress σ{sub θ}, and cladding material temperature affect cladding ductility? Is hydride-related cladding embrittlement a material degradation mechanism that affects the structural integrity of PWR SNF rods under pinching loading? What are the controlling parameters for deformation demands in PWR SNF rods under pinching loading? Is the pinching load amplitude an important parameter that controls the probability of pinching failure? How do parameters like pellet-cladding gap or fuel pellet degradation affect the pinching loading capacity of PWR SNF rods? Statistical modelling of PWR SNF rod conditions in U.S. inventory: A pseudo rod database was generated based on EIA database. Analysis of train accident scenarios: a probability density function (pdf) of a pinching load distribution P{sub 9m} of a 9-m cask drop is linearly scaled to generate a more representative: distribution P{sub AS}. The new distribution considers train collision velocities, cask drops for on- and off-bridge accidents, and cask impact orientation. LHS generates 3,000 rod pinching scenarios based on nine independent geometrical, material, and loading parameters. Probabilistic FE Analysis: The LHS realizations are used to generate 3,000 FE models. The FE analysis is conducted for P{sub 9m}, and P{sub AS}. The workflow is automatized. The OSC is computed with regression and interpolation models, as a function of the hydrogen content C{sub H}, vacuum drying peak σ{sub θ}, and SNF temperature T{sub FR}: The regression data was obtained in RCTs on PWR cladding. The OSD is derived from the cladding deformations in the FE models. Cladding failure was identified via a comparison of OSC and OSD. A 300 and a 30 years-of-dry-storage duration is investigated: The dry-storage period affects the expected SNF temperature and consequently, the OSC and the OSD. The material temperature LHS population is adapted to account for a SNF temperature variation. The probability of hydride-related, ductile cladding failure during a severe transportation accident is not significantly affected by the pinching load amplitude: The probability is slightly reduced from a range of less than 0.0333% to 0.1667% for P{sub 9m}, to a range of less than 0.0333% to 0.1333% for P{sub AS}, depending on the vacuum drying peak cladding hoop stress modelling assumptions. Further, the SNF dry-storage duration does not significantly affect the probability of ductile pinching failure. Based on available train accidents data, the pinching failure probability for a 5,000 km long transportation route ranges from less than 1.62 E-7 to a value of 6.47 E-7, depending on the vacuum drying peak cladding hoop stress modelling assumptions. The pinching load amplitude is not a controlling parameter for the probability of ductile cladding pinching failure, and does not control the cladding deformations once pellet-cladding contact is established, The pellet-cladding gap width, however, plays a key role. Fuel pellet degradation only affects the deformation capacity, if the pellet stiffness is significantly reduced (more than 90%). Only highly embrittled cladding is likely to fail. The deformation demands on the cladding under pinching loading are small.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Multiple RANS Scheme for the CFD Simulation of PWR

Though best-estimate system codes have been well established, the assumptions for the simplification of the conservation equations will be invalid when 3D effects play an important role in the object. The inner domain of PWR can create high 3D effects because of the disturbance of mixing vanes and the complex distribution of flow resistance. With the finest mesh to study the flow and heat transfer process of a nuclear reactor, a CFD simulation can give a perfect understanding of the physical processes in a whole reactor. However, more than 10 thousands of subchannels exist in a typical pressurized water reactor (PWR). These subchannels are narrow, comprising complex spacer grids and mixing vanes. The need of both the computational resources and the simulation time is great, which is not good for the simulation efficiency. However, the efficiency is critical for the CFD simulation of a PWR. It is meaningless if the simulation does not end. To improve the efficiency, one approach is to well consider the turbulence model used in a CFD simulation, because some models are complex and require great computing resources and time. On the contrary, some models are simple, but are also suitable to reflect the physical status in a PWR. Furthermore, the effects of the typical structures of PWR should also be clarified. They are closely related to the flow mechanism. The CFD simulation may be simplified by such effects. The objective of this research is to study the flow characteristics affected by the turbulence models and typical structures. The authors then design a scheme based on the regularities of the flow characteristics to improve the efficiency of the CFD simulation for PWR. By this research, we found: 1) The mixing vanes, a typical structure of a PWR, make the anisotropic eddy viscosity cannot be ignored for the region with short distance downstream of this structure; 2) Outside of this region, the RANS models, regardless of whether they are isotropic or anisotropic, can obtain similar results. To improve the simulation efficiency, a CFD scheme is designed by the hybrid application of multiple RANS models (MRANS). In this scheme, the major domain with low anisotropic viscosity uses the turbulence model which requires the least time. The difference is just little in the comparison with a simulation using the anisotropic RANS model for the whole computational domain. The MRANS scheme thus will be significant to improve the efficiency of the CFD simulation for a whole PWR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Development and Demonstration of a Reduced-Order PWR Power Dispatch Simulator

This report describes the development, modeling, and results of a reduced-order thermal power dispatch pressurized water reactor (RO-TPD-PWR) power plant simulator that incorporates coupled electrical and thermal power dispatch to an industrial process. The simulator is built on models that have been developed in previous work, including a Rancor Microworld model previously developed by INL and the University of Idaho [Ref] and reactor core and secondary system models that have been published previously [Ref]. The simulator also includes a scalable model of a high temperature electrolysis (HTE) system that has been developed at INL. A particular benefit of the Rancor Microworld model is that it is compatible with a human-system interface (HMI) that has already been developed by the Human System Simulation Laboratory (HSSL) at INL that can emulate the control room of a PWR for realistic human-in-the-loop studies with mock nuclear power plant operating procedures. The results of these human-in-the-loop tests will be compared with similar tests using a commercial full-scope generic PWR (GPWR) simulator that has been modified to include thermal power dispatch, referred to as the TPD-GPWR [ , ]. The simplicity of the RO-TPD-PWR helps to better understand and verify the results of the TPD-GPWR, which employs >1,000 variables and has much greater complexity. A second key purpose of the RO-TPD-PWR simulator is that its relative simplicity allows it scale between lab-size equipment (~100 kW) and full-scale nuclear power plants (~1 GW) by adjusting only a few parameters in the models, such as fluid masses and heat exchanger areas. The RO-TPD-PWR simulator will be used to guide the controls of lab tests, so that equipment in those tests, such as the Thermal Energy Distribution System (TEDS) and pilot-scale high temperature electrolysis (HTE) units at scales of 50-250 kW, may operate in a manner that is relevant to full-scale PWR operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Prediction of countercurrent flow limitation and its uncertainty in a PWR hot leg

In previous studies, we did small-scale air-water experiments for the hot leg and pressurizer surge line in a PWR and three-dimensional (3D) numerical simulations for PWR conditions to evaluate countercurrent flow limitation (CCFL) there, and we developed a one-dimensional (1D) prediction method to generalize prediction of CCFL in a piping system such as the hot leg and pressurizer surge line. In this study, we evaluated computation errors of the 1D prediction method in comparison with CCFL data in the PWR hot leg models with the diameter D and length L of D = 50-650 mm and L/D = 4.5-9.5, and then evaluated uncertainties of the 1D prediction method in comparison with 3D numerical simulations for PWR conditions with D = 750 mm and L/D = 8.6. We found the computation errors were within the difference of dC = ±0.05 for the CCFL constant of C = 0.63 in the correlation with the Wallis parameters, but the computation errors include experimental errors and they are not uncertainties for the PWR conditions. The total uncertainty of the 1D computations for the actual PWR hot leg (including effects of the diameter and fluid properties) was estimated to be dC = ±0.035. The CCFL correlation and its uncertainty of dC = ±0.035 derived from the 1D prediction method were very close to those previously obtained from CCFL data and 3D simulations (dC = ±0.03). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Comparison of selective oxidation in Ni-based alloys exposed to PWR primary water and Rhines Pack environments

A mechanistic understanding of corrosion and oxidation processes is crucial to ensure long-term resistance to stress corrosion cracking of Ni-based structural alloys in PWR primary and secondary systems. Aqueous corrosion of complex Ni-based alloys under PWR water conditions as well as gaseous oxidation of low-solute Ni binary alloys at high temperatures (>800 °C) have been studied extensively. However, a significant knowledge gap exists between response of Ni-based alloys in aqueous and gaseous environmental conditions. Selective oxidation at different temperatures was investigated in a series of gaseous Rhines Pack experiments for a high-purity Ni-5Cr binary alloy and a commercial Ni-18Cr-9Fe Alloy 600. Focused ion beam milling enabled the extraction of site-specific specimens containing high-energy grain boundaries. Analytical transmission electron microscopy was employed to analyze the microstructure and chemical composition of the resulting localized oxidation. The oxidation mechanism in the gaseous Rhines Pack atmosphere at 420 °C included the formation of a protective Cr-oxide cap above grain boundaries in all cases. At higher temperatures (600 and 800 °C), penetrative intergranular and transgranular internal oxidation was observed. Direct comparisons are made to corrosion response for these same materials in simulated PWR primary water at 320-360 °C. These results reveal a temperature dependence for gaseous oxidation and indicate mechanistic differences between aqueous and gaseous degradation in Ni-Cr alloys. These results imply that selective intergranular corrosion in PWR water is governed by processes beyond mere oxidation. Consideration of complex corrosion/dissolution processes at surfaces and in open cracks is essential for the better understanding of stress corrosion cracking mechanisms in PWR primary water.

Rhines Pack, Alloy 600, Ni-base alloys, Internal O↗
DOE OSTI2015

Development of a multi-purpose container system for the storage, transport and disposal of PWR spent fuel in the UK - 15466

Radioactive Waste Management Limited (RWM) is a wholly owned subsidiary of UK's Nuclear Decommissioning Authority, and is responsible for implementing UK Government policy on geological disposal of higher activity radioactive waste. An aspect of the generic work programme that underpins the current generic disposal concepts includes the development of the associated disposal containers. To date RWM has focused on the development of a range of disposal container designs that fulfil the requirements for transport and disposal in a Geological Disposal Facility (GDF) for legacy high heat generating wastes, one of which includes legacy Pressurised Water Reactor (PWR) spent nuclear fuel (SF). In line with a number of European countries exploring deep geological disposal, the conceptual disposal container solution developed for this waste type has a capacity of 4 PWR fuel assemblies. Since 2013, the feasibility of developing an alternative higher capacity disposal container design based around the Multi-Purpose Container (MPC) concept has been explored. In general terms, MPCs are containers that are designed to meet requirements for safe containment of radioactive waste during storage, transport and disposal. The fundamental design philosophy of the MPC is that once the waste is packaged in an MPC, it would not need to be handled again. That is, after a period of initial cooling in a fuel pond, the SF could be packaged for long term dry storage and would not require direct handling for its eventual transport to, and disposal in, a GDF. The novel feature of the MPC system design considered is the inclusion of an additional overpack specifically for the final deep geological disposal phase. The conceptual MPC and its associated storage, transfer, transport and thick walled carbon steel disposal overpack have been developed to a conceptual level of design detail to demonstrate concept feasibility, with the final design solution confirming a MPC system with a capacity of 12 PWR SF assembles. This represents the highest capacity PWR MPC system that complies with the mass and rail gauge limits of the UK W6a rail gauge. Such a capacity realises a contents loading three times that of the standard PWR disposal container and presented technical challenges from both a structural performance and thermal perspective; necessitating the need for the inclusion of features to enhance heat transfer from the central fuel assemblies within the MPC and novel lid features on the transport container overpack to maintain structural integrity and containment during the hypothetical accident conditions of transport tests. In developing the MPC system conceptual design it is recognised that, depending upon the safety functions of the MPC disposal overpack, the choice of carbon steel material may not be suitable for all geological environments. Consequently, the MPC system design and its materials of construction will require further evaluation once detailed knowledge of the disposal concept to be implemented, its safety functions and geological environment are known. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

An industrial application of the burnup credit including actinides and fission products for PWR UO{sub 2} used fuel transportation - 14590

Criticality safety analyses related to the transportation of PWR uranium oxide (UO{sub 2}) used fuel assemblies are usually performed under the fresh fuel assumption. This leads to huge safety margins, as the negative reactivity worth due to the irradiation of the assemblies is not taken into account. Since 1987, AREVA TN has been using a limited burnup credit method based on the sole consideration of major actinides and the use of a uniform axial burnup profile corresponding to the mean burnup at the 50 least-irradiated centimetres of the assembly. This method is quite limited today for some applications, due mainly to the increase of the fuel enrichment in PWR UO{sub 2} fuel assemblies. The strategy of AREVA TN to deal with high enriched PWR UO{sub 2} fuel assemblies and also to limit the increase of the neutron poison content in the new baskets' designs is to take benefit from the reactivity reserve, which can be gained by considering main fission products and less penalizing axial burnup profiles instead of uniform axial burnup profiles. The Burnup Credit (BUC) calculation route for PWR UO{sub 2} used fuel is based on the connection of the depletion code DARWIN (developed by the CEA) and the Criticality Safety Package CRISTAL V1 (developed by the CEA and the IRSN in partnership with AREVA and EDF). French BUC experimental programs have been also separately performed in Cadarache and in Valduc in order to validate respectively the APOLLO2-DARWIN2 depletion code and the CRISTAL V1 Criticality Safety package. This paper presents the BUC methodology including actinides and fission products but also the approach used for the criticality safety assessment of representative transport cask loaded with PWR UO{sub 2} used fuel assemblies approved for the first time by the French Safety Competent Authority. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 1: Pin and lattice

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. This paper focuses on analysis of the predictions for pin and lattice designs representative of HB/HE core reloads from the Westinghouse advanced lattice code PARAGON2, the Monte-Carlo code SERPENT2 and MPACT, the deterministic neutronic component of VERA. A companion paper presented at this conference analyzes results obtained with the Westinghouse ANC9 core simulator, using lattice data provided by PARAGON2 with the NEXUS cross-section methodology, and MPACT for 2D and 3D core configurations incorporating HB/HE fuel reloads to achieve cycle length extension to 24-month in high-power density PWRs. The results of the pin and lattice benchmark presented in this paper show that the neutronic predictions from PARAGON2 for HB/HE fueled PWRs are in remarkable agreement with predictions from both SERPENT2 and MPACT including for depletion to high fuel burnup values, and employing a set of challenging lattices which feature combined use of burnable absorbers to obtain the required reactivity hold-down to support PWR transition to HB/HE fuel and 24-month operational cycles. As corroborated by this analysis and the analysis contained in the companion paper, the Westinghouse NEXUS-based PWR core analysis package can be confidently applied by industry for design analysis of PWR core reloads adopting HB/HE fuel and properly support the industry commercialization plans for in-reactor deployment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of Advanced T-OZON{sup TM} for PWR Decommissioning (2) - 18170

The number of aging nuclear power plants (NPPs) slated for closure around the world has increased. The majority of them are PWRs. Minimization of radiation exposure for PWR decommissioning workers is one of the key factors in decommissioning in order to apply the as low as reasonably achievable (ALARA) principle. To this end, it is desirable to establish a chemical decontamination method to reduce the exposure of PWR decommissioning workers. Toshiba developed T-OZON{sup TM}, a chemical decontamination method for BWRs [1, 2, 3]. To date, the method has been applied more than 90 times in Japanese BWR plants. The reduction in radiation exposure of workers achieved by applying T-OZON{sup TM} is estimated at about 300-2700 man mSv [4]. A major feature of T-OZON{sup TM} is the reduced amount of radioactive waste produced when compared to other decontamination processes. Oxide films on PWR wetted surfaces are harder to decontaminate than those in a BWR due to their chemical composition. To apply T-OZON{sup TM} to PWR oxide films, it is necessary to modify the process conditions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Final Report on CFD and Thermal-Mechanical Stress Analysis of PWR Surge Line under Transient Condition Thermal Stratification and an Evolutionary Cyclic Plasticity Based Transformative Fatigue Evaluation Approach without Using S~N Curve: Rev. 1

This report presents an evolutionary cyclic plasticity based transformative approach for fatigue evaluation of safety critical components. This approach is generic which can be used not only for the present generation light water reactor components (which is our focus) but also for advanced reactor components. The approach is based on the fundamental concept of time-evolution of progressive fatigue damage rather than the end-of-life data based conventional S~N curve approaches. Series of fatigue tests on 316 stainless steel uniaxial specimens were conducted under strain-controlled and stress-controlled cyclic loading in two different environment such as in-air at 300 °C and PWR coolant water condition at 300 °C. Material models are developed to capture the uniaxial test data and are subsequently programmed into commercially available ABAQUS code to transform the uniaxial material behavior to multiaxial loading domain. The developed evolutionary cyclic-plasticity approach is verified by both analytical and 3D-FE modeling of uniaxial fatigue test specimens. Results show that the developed material models can capture the time-dependence of material hardening and softening with great accuracy. In addition, results show that the material models not only can capture the material behavior under constant amplitude loading but also can capture the load-sequence effect under variable and random amplitude loading. The ANL developed approach is one of its kind, which shows first in the fatigue research community that fatigue evaluation of a component (in our case verified for laboratory scale fatigue test specimens with great accuracy) can be performed for its entire fatigue life, including the hardening and softening behavior, without using a conventional S~N curve based approach. Additionally results are presented related to fatigue evaluation of real reactor components such as of a pressurized water reactor (PWR) surge line (SL), under upset transient conditions. SLs of PWR experience complex thermal stratification due to the mixing of differential temperature flow from hot leg and pressurizer. A detailed parametric study of thermal stratification in SL is performed through both standalone computational fluid dynamics (CFD) analysis and fluid-structure-interaction based coupled CFD and heat transfer analysis. Results show that the mass flow rate in SL substantially affects the extent of its thermal stratification. The resulting thermal boundaries conditioned are used as inputs for thermal-mechanical stress analysis and component level fatigue evaluation. Fatigue evaluation of SL pipe under example loading conditions was performed using the above-mentioned evolutionary-plasticity based approach and ASME code (for in-air life) and NUREG-6909 (for PWR environment life) based approaches. Results show that the evolutionary plasticity based approach estimate substantially different life compared to the conventional ASME based approach. To show the entirety of the proposed evolutionary fatigue-modeling framework, this report includes the results presented in semi-annual March report (ANL/LWRS-17/01). Sections 2-5 of March report are directly included as Sections 2-5 in this report, without any significant changes. However, Sections 6-9 of this report present the work performed during March-September of 2017. Whereas, Sections 1 and 10 present the organization of this report and the summary and future work, respectively. This report (ANL/LWRS-17/03 Rev. 1) is a slightly revised version of the earlier published report (ANL/LWRS-17/03). In this revised version, some of the results and associated discussion have been corrected/updated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Microstructural Characterization of the Second High Fluence Baffle-Former Bolt Retrieved from a Westinghouse Two-loop Downflow Type PWR

As one of the pressurized water reactor (PWR) internal components, baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. In support of evaluating long-term operational performance of materials used in core internal components, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, Materials Research Pathway (MRP) has harvested two high fluence BFBs from a commercial Westinghouse two-loop downflow type PWR. The two bolts of interest, i.e. bolts # 4412 and 4416, were withdrawn from service in 2011 as part of a preventative replacement plan. No identification of cracking or potential damage was found for these bolts during their removal in 2011. However, the bolts required a lower torque for removal from the baffle structure than the original torque specified during installation. Irradiation displacement damage levels in the bolts range from 15 to 41 displacements per atom. The goal of this project is to perform detailed microstructural and mechanical property characterization of BFBs following in-service exposures. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long-term irradiation behavior and compared to results obtained in high flux experimental reactor conditions. In this report, we present our latest study in FY22 on microstructural characterizations of the second high fluence baffle-former bolt, i.e., bolt # 4412. Analytical electron microscopy and atom probe tomography characterization were performed. The radiation-induced defects in the material add to the large wealth of knowledge for neutron-induced defects in 304/316 grades of stainless steels, specifically for radiation-induced precipitation after high fluence commercial PWR irradiation. The main findings are summarized as follows: 1) The cavity size was considerably larger in the bolt thread section than in the bolt head, with the bolt thread section having a bimodal distribution of cavities greater than ~6 nm in diameter and less than ~3 nm in diameter. The bolt head only had the small-sized cavities. In addition, there was a denuded zone of large cavities near grain boundaries in the thread section of the bolt. 2) Radiation-induced precipitation in the BFB #4412 was highly complex, with the volume fraction, size, and number density of Ni/Si and Cu-rich precipitates depending strongly on the radiation temperature/dose. In many cases, co-precipitates of adjoined clusters were found with Ni/Si-rich precipitates sandwiched between Cu-rich clusters and Mo/Cr/P-rich clusters. 3) Solute segregation out of solution was highest for most solutes in the thread section of the bolt #4412 with the exception of Cu, which experienced more separation out of solution into Curich clusters in the bolt head section. This highlights the difference in the mechanisms for precipitation of Ni/Si clusters, which have the Ni 3 Si phase composition, and precipitation of Cu-rich clusters. 4) There appear to be multiple simultaneous influences that affect the microstructural variation along the length of the bolt that overcomes the ~2X difference in irradiation dose between the bolt head and the bolt thread. The irradiation temperature, thermal/fast neutron ratio variation, potential strain gradient, and exposure to PWR coolant water that each section of the bolt sees may have more influence on the microstructural evolution than the total irradiation dose. The bolt thread and shank, with higher temperature, higher relative fast neutron flux, higher strain, and exposure to coolant but lower dose, underwent more enhanced cavity formation, precipitation, and solute segregation than the bolt head section.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Interface capturing simulations of bubble population effects in PWR subchannels

As the computational power of high-performance computing (HPC) facilities grows, so too does the feasibility of using first principle based simulation to study turbulent two-phase flows within complex pressurized water reactor (PWR) geometries. Direct numerical simulation (DNS), integrated with an interface capturing method, allows for the collection of high-fidelity numerical data using advanced analysis techniques. The research presented here employs the massively parallel, finite-element based, unstructured mesh code, PHASTA, to simulate a set of two-phase bubbly flows through PWR subchannel geometries including auxiliary structures (spacer grids and mixing vanes). The main objective of the presented work is to analyze bubble dynamics and turbulence interactions at varying bubble concentrations to support the development of advanced two-phase flow closure models. Turbulent two-phase flows in PWR subchannels were simulated at hydraulic Reynolds numbers of 81,000 with bubble concentrations of 3%–15% by gas volume fraction (768–3928 resolved bubbles, respectively) and compared against a 1% void fraction case (262 bubbles) that had been previously simulated. The finite element mesh utilized for the study at higher bubble concentrations was composed of 1.55 billion elements, compared to the previous study which employed 1.11 billion elements, ensuring all turbulence scales and individual bubbles within the flow are fully resolved. For each case, the resolved initial bubble size was 0.65 mm in diameter (resolved with 25 grid points across the diameter). The simulations were analyzed to find flow features such as the mean velocity profile, bubble relative velocity and the effect of the bubbles on the turbulent conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Westinghouse advanced fuel management strategies leveraging high enrichment and high burnup fuel to optimize PWR economics

In the US, nuclear power plants have been focused on increasing electricity generation for over three decades via longer operating cycles, reduced number of outages and duration, and the implementation of power up-rates. The potential for zero-carbon emission credits and other carbon-reduction initiatives further makes nuclear power utilities to focus on increasing generation capability. Longer operating cycles provide increased energy generation and enable economic savings to nuclear utilities when the increase in fuel loading required to achieve the extended cycle energy target and resulting increase in total fuel cost is offset by the savings from outage avoidance and reduced replacement power costs. For the current licensed limits of 5 w/o {sup 235}U enrichment and 62 GWd/tU peak pin burnup, higher power density PWR plants, which are prevalent in the PWR fleet, require an excessively penalizing fraction of feed fuel assemblies to operate for the extended cycle duration, with inevitably poor fuel use and fuel cycle economics that cannot be counterbalanced by the savings on outage cost avoidance. To overcome this barrier, the nuclear industry is pursuing development of high enrichment/high burnup fuel technology ({sup 235}U enrichment up to 8 w/o and peak pin burnup up to 75 GWd/tU) which increases fuel energy generation capability and can enable high power density plants to achieve positive economics on 24-month cycle of operation. This paper presents advanced fuel management strategies developed by Westinghouse to enable transition of PWR reactors from 18-month to 24-month cycle of operation, with and without a core thermal power uprate, which are then used as basis to assess the economics of a direct transition from 18 to 24-month cycles with high enrichment/high burnup fuel. The results show that these fuel management strategies, especially if coupled with thermal power up-rates, present the opportunity for substantial economic benefits to utilities. The systematic review of fuel management strategies presented can provide an illustrative reference for decision makers as they consider how to maximize nuclear energy generation economically. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗