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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Thermal-Mechanical Stress Analysis of PWR Pressure Vessel and Nozzles under Grid Load-Following Mode: Interim Report on the Effect of Cyclic Hardening Material Properties and Pre-existing Cracks on Stress Analysis Results

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.

42 ENGINEERING↗
DOE OSTI2016

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Stress Corrosion Crack Initiation of Cold-Worked Alloy 600 and Alloy 690 in PWR Primary Water

This is an annual report summarizing research on stress corrosion crack initiation in alloy 600 and alloy 690 in simulated pressurized water reactor primary water environments. A key aspect of the work is developing an understanding of the mechanisms and controlling factors for producing initiated cracks in these materials. This work covers research that explores and compares the initiation response between alloy 600 and alloy 690.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to a particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of Gd{sub 2}O{sub 3} burnable poison on the measurement of fresh pressurized water reactor fuel. To empirically determine the response function over the range of historical and future use we have considered enrichments up to 5 wt% {sup 235}U/{sup tot}U and Gd weight fractions of up to 10 % Gd/UO{sub 2}. Parameterized correction factors are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

STRESS CORROSION CRACK GROWTH RESPONSE FOR ALLOY 152/52 DISSIMILAR METAL WELDS IN PWR PRIMARY WATER

As part of ongoing research into primary water stress corrosion cracking (PWSCC) susceptibility of alloy 690 and its welds, SCC tests have been conducted on alloy 152/52 dissimilar metal (DM) welds with cracks positioned with the goal to assess weld dilution and fusion line effects on SCC susceptibility. No increased crack growth rate was found when evaluating a 20% Cr dilution zone in alloy 152M joined to carbon steel (CS) that had not undergone a post-weld heat treatment (PWHT). However, high SCC crack growth rates were observed when the crack reached the fusion line of that material where it propagated both on the fusion line and in the heat affected zone (HAZ) of the carbon steel. Crack surface and crack profile examinations of the specimen revealed that cracking in the weld region was transgranular (TG) with weld grain boundaries not aligned with the geometric crack growth plane of the specimen. The application of a typical pressure vessel PWHT on a second set of alloy 152/52 – carbon steel DM weld specimens was found to eliminate the high SCC susceptibility in the fusion line and carbon steel HAZ regions. PWSCC tests were also performed on alloy 152-304SS DM weld specimens. Constant K crack growth rates did not exceed 5x10-9 mm/s in this material with post-test examinations revealing cracking primarily on the fusion line and slightly into the 304SS HAZ.

alloy 152↗
DOE OSTI2012

Comparison of homogenized and enhanced diffusion solutions of model PWR problems

Model problem comparisons in slab geometry are made between two forms of homogenized diffusion theory and enhanced diffusion theory. The pin-cell discontinuity factors for homogenized diffusion calculations are derived from homogenized variational nodal P1 response matrices and from standard finite differencing. Enhanced diffusion theory consists of applying quasi-reflected interface conditions to reduce variational nodal Pn response matrices to one degree of freedom per interface, without homogenization within the cell. As expected both homogenized diffusion methods preserve reaction rates exactly if the discontinuity factors are derived from the P 11 reference solutions. If no reference lattice solution is available, discontinuity factors may be approximated from single cells with reflected boundary conditions; the computational effort is then comparable to calculating the enhanced diffusion response matrices. In this situation enhanced diffusion theory gives the most accurate results and finite difference discontinuity factors the least accurate. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

PWR cores with silicon carbide cladding

The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Status of verification and validation of AREVA's ARCADIA{sup R} code system for PWR applications

In March 2010 the submittal of Topical Reports for ARCADIA{sup R} and COBRA-FLX, the thermal-hydraulic module of ARCADIA{sup R}, to the U.S. Nuclear Regulatory Commission (NRC) concluded a major step in the development of AREVA's new code system for core design and safety analyses. This submittal was dedicated to the application of the code system to uranium fuel in pressurized water reactors. The submitted information comprised results for plants operated in the US (France)) and Germany and provided uncertainties for in-core measuring systems with traveling in-core detectors and for the aero-ball system of the EPR. A reduction of the uncertainties in the prediction of F{sub AH} and F{sub Q} of > 1 % (absolute) was derived compared to the current code systems. This paper extents the verification and validation base for uranium based fuel and demonstrates the basic capabilities of ARCADIA{sup R} of describing MOX. The achieved status of verification and validation is described in detail. All applications followed the same standard without any specific calibration. The paper gives also insight in the new capability of 3D full core steady-state and transient pin-by-pin/sub-channel-by-sub-channel calculations and the opportunities offered by this feature. The gain of margins with increasing detail of the representation is outlined. Currently, the strategies for worldwide implementation of ARCADIA{sup R} are developed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

PWR internal flow modeling with fuel assemblies details

This study is an example of a massive parallel computing of the coolant flow in a nuclear reactor. It resolves the flow velocities in each assembly on pin level and predicts the flow distribution in complex geometries such as the lower and upper reactor plenums. The size of the developed model (1.035 billion cells) required the runs to be executed on the NCCS clusters (www.nccs.gov). STAR-CCM+ code (www.ed-adapco.com) was installed on two clusters: JAGUARXT5 and FROST, both of which were capable of executing this model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Surveillance of PLUS7{sup TM} fuel for PWR nuclear power plant

The surveillance program on the advanced nuclear fuel of PLUS{sup TM} developed for Optimized Power Reactors of 1000 MWe (OPR1000s) and Advanced Power Reactors of 1400 MWe (APR1400s) in Korea was completed in the early of 2011. This fuel had been jointly developed through the extensive out-of-pile tests with Westinghouse for three years since 1999. The irradiation tests for the in-reactor verification using four lead test assemblies (LTAs) had been started in Ulchin unit 3 in 2002. During the overhaul period after each irradiation test, the eight (8) burnup-dependent parameters were measured without disassembling using the precise measurement systems in pool-side. After three cycle irradiations, one test assembly was disassembled and the rod-wise inspection on twenty rods was performed. During this stage, five (5) parameters were measured and evaluated. Among these twenty rods, ten rods including skeleton were sent to hot-cell test facility for further detailed examination and are currently being examined. After in-reactor verifications during two cycles, this fuel was commercially supplied to eight (8) OPR1000s sequentially. Currently all eight (8) OPR1000s were replaced with this fuel. In addition, this fuel is going to be supplied to four (4) APR1400s being constructed in Braka, UAE as well as four(4) OPR1000s and four(4) APR1400s being constructed in Korea. In the meanwhile, the surveillance program for the commercially supplied fuel has been launched to confirm growth, creep, corrosion and deformation, etc. obtained during LTA irradiation. Four (4) limiting fuel assemblies, that is, two (2) assemblies to be discharged after 2 cycle irradiations and the other two (2) after 3 cycle irradiations were selected for this surveillance program. Irradiation data of commercially supplied fuels are compared and confirmed to LTA irradiation performance behaviors on this paper. Among the eight (8) burnup-dependent parameters, the interesting ones were irradiation-induced growths of grid width, rod and assembly lengths while cladding oxide thickness, rod creep and swelling followed the LTA results, and assembly bow/twist and rod bow were well below the design limits. The rod growths were on the lower bound of database and contributed to assembly growth. It was evaluated that these results were attributed to PLUS{sup TM} design characteristics which adopted Inconel top and bottom grids and conformal-shaped mid grid spring. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Development of a new flux map processing code for moveable detector system in PWR

This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

On-line measurements of RuO{sub 4} during a PWR severe accident

After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

On-line fission products measurements during a PWR severe accident: the French DECA-PF project

Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, funded in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied: - Gamma spectrometry, with an industrial objective to build a prototype aimed at improving the capacity of the present radiation monitoring system, - Gas chromatography, for the quantification of the fission gases (Xe, Kr) as well as potential carbon oxides produced in case of Molten Corium Concrete Interaction, - Optical absorption spectroscopy, the objective of this most innovative technique being to quantify the tetra-oxide of ruthenium, which could be produced in case of lower head failure, and the gaseous forms of iodine (molecular and organic) released in the environment. A global description and the present status of this project is presented, focusing on the Source Term establishment at the outlet stack of the sand bed filters and on the perspectives of implementation of the on-line gamma spectrometry equipment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

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

Enhancement of Nuclear Engineering Technology Degree with a Web-Based PWR Simulator

Reactor plant simulators are used in the nuclear industry for initial and continuous training of reactor operators and senior reactor operators. The use of the generic pressurized water reactor (GPWR) plant simulator is valuable for the nuclear industry as it will ensure graduates of the Bachelor of Science in Nuclear Engineering Technology (BSNET) program will be better trained employees for the nuclear industry and address forecasted workforce shortages in the nuclear industry. By adopting simulation for education, a proven teaching and training method applied extensively in the plant environment, work quality among entrants to the nuclear power workforce will be improved, elevating performance in the field. This includes a positive influence on safety which is paramount to retaining public support of this clean energy source. Much of the research on the use of simulations within courses has argued that simulations are good tools to help students learn key concepts, terminology, and theories within courses. The GPWR simulator provides an opportunity for nuclear engineering technology students to improve their understanding of theoretical concepts associated with nuclear power plants. This hands-on experience will ensure graduates of the BSNET program will be better trained employees for the nuclear industry and address forecasted workforce shortages in the nuclear industry by providing a real-world experience in plant operations. The GPWR simulator will be utilized in five nuclear courses to begin, while providing opportunities for future expansion of the curriculum. All students graduating with a BSNET degree will interact with the simulator, experience better preparing them for further simulator training provided throughout the industry for licensure of operators. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗