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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Radiochemical Characterization of PWR Control Rods - 19030

NRG has, in cooperation with the German partner Hoefer and Bechtel, performed a radiochemical characterization of the control rods from a German Pressurized Water Reactor. To perform the cutting and sampling of the spent control rods at the reactor site NRG has developed a dedicated, remotely operated tool. This tool enabled the sampling of both the stainless steel tube and the absorber material from the upper, middle and lower parts of selected rods. The samples were transported from the reactor site to NRG's laboratories at Petten, the Netherlands, were they were analysed chemically and radiologically. Due to the different chemical constitution of the control rod tubes and the absorber, the procedures for the chemical separation of the different compounds, prior to the radiological characterization, varied. The route for the stainless steel tube samples turned out more complex than that for the absorber samples: there are more nuclides to be determined, and therefore more treatment steps and component separations were necessary. Subsequently to the component separations, the radiochemical characterization was performed using alpha and gamma spectrometry and the Liquid Scintillation Counting (LSC) technique. The results of the measurements in combination with simulations performed with a reactor physics code provided for a better understanding about the distribution of radionuclides in absorber rods, both in the rod tubes and the absorber core. In addition, the results of the measurements can be used to validate reactor physics codes. (authors)

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

PWR Station Blackout and External Events Analysis for LWRS/RISMC Project

In the framework of the US-DOE Light Water Reactor Sustainability/ Risk-Informed Safety Margin Characterization (LWRS/RISMC) project, methods, data and tools are being developed with the ultimate scope of improving the safety and operation of the current US nuclear fleet. A set of Industrial Applications (IA) concerning the most relevant safety issues is defined for applying and testing the developed tool kits. The ultimate scope of IA no. 2 is a realistic simulation of natural external hazards that impose threat to a Nuclear Power Plant (NPP). This scope requires the development of a methodology and of a qualified set of tools able to perform advanced risk-informed safety analysis. In particular the methodology should be able to combine results from seismic, flooding and thermal-hydraulic (TH) deterministic calculations with dynamic Probabilistic Risk Assessment (PRA). A so-called Risk Informed Margin Management (RIMM) approach is being used to represent meaningful (i.e., realistic facility representation and simulation) event scenarios and consequences by linking probabilistic simulations to external events physics. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Modeling PWR systems with a unique framework for hybrid energy systems

Hybrid energy system (HES) is used to utilize the wind, solar, nuclear and other energy sources. Within the study and design of a nuclear power plant and HES, extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including unanticipated transients without scram and hypothetical accidents. A modelling tool that is capable of solving governing equations of mass, momentum and energy for two-phase application is necessary to model the coolant systems of HES. Therefore, this study shows the development of a new coupling tool that is capable of modelling several components of HES and experimental validation of the model. For the initial development phase, the primary and secondary coolant systems (which use several components similar to the components of HES) are modelled by using this new tool. Code-to-code benchmark is also used for the performance evaluation of the new tool. It is observed that there is a good agreement between the new tool and RELAP5 predictions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Towards a Neutronic and Thermal-Hydraulic Safety Analysis of PWR Fuel Bow with SIMULATE5 and COBRA-TF

In the 1990's, collective fuel assembly bow was observed in the Ringhals PWRs in Sweden. Back then fuel vendors reinforced the fuel assemblies' dashpot, decreased hold-down spring forces, increased guide tube wall thickness, and introduced guide tube materials with higher creep resistance in order to improve dimensional stability. Nevertheless, twenty years later in 2010, the same kind of collective bow with S and C shape patterns started to appear in most of Germany's PWRs. At this time the SIMULATE5 (S5) core simulator with its delta-gap bow model (DGA) which was originally developed for safety analysis at Ringhals, could be used as an off-the-shelf solution to assess the impact of bow on Fq and FH. However, these two quantities are of secondary importance in German PWRs in normal operation because the most restricting licensing criteria set limits on DNBRmin values. For this purpose, a method was developed to extract S5 pin powers from DGA calculations and automatically generate COBRA-TF input files (CTF) to analyze DNBR values in the cores. In the following two sections, it is shown how the S5 model for DGA calculations has been verified with independent MCNP6 calculations in order to check its predictions on pin power changes for delta gaps up to 1 cm in size and how the existing licensed DNBR calculation methodology has been updated to bowed fuel assembly situations with the help of CTF. Finally, the applicability of the Doroschuk CHF correlation, which is the licensing standard in most German PWRs, is discussed. For the fuel bow patterns considered, it is assumed that only the gap between fuel assemblies change and that the pitch between fuel rods inside the assemblies remains largely unchanged. This approximation can be justified by the visual inspections conducted at the plants during the outages. Every calculation of the pin-power impact of delta-gaps in the core under operating conditions is only as good as the knowledge of the bow pattern. Measurements of fuel assembly bow at cold conditions have been made for some German PWRs for all the fuel assemblies in the core during plant outages. Together with methods developed by the fuel assembly vendors, a best-estimate bow pattern during operation could be reconstructed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Mapping Turbulent Flow around a Rod Bundle Induced by a 5 x 5 PWR Mixing Vane Grid using Particle Tracking Velocimetry

This study presents an experimental visualization technique developed for the measurement of the lateral liquid velocity components within a vertical up-flow 5 x 5 rod bundle. The experimental technique uses refractive index matching (RIM) and time resolved particle tracking velocimetry (PTV) to obtain full field measurements at different axial locations. Relevant features of the technique are first, the multi-scale capability, which allows measurements either on a single subchannel or measurements to include the full bundle accounting for the 25 rods and corresponding liquid subchannels. Secondly, the use of FEP tubes to match the refractive index of water allows a uniform illumination which in turn reduces the velocity measurements uncertainty, and third, the use of a high energy continuous laser allows for the temporal evaluation of fluctuating structures within the bundle. This study was conducted with the aim to quantify the effect of the mixing vanes on the development of the liquid turbulence and to test the motivation feasibility of the technique to serve as a benchmark for advanced CFD simulations. In this context, there have been multiple experimental studies with the objective to improve the database for CFD benchmarking purposes. Several experimental techniques have been used to this end; for instance, laser doppler velocimetry (LDV) is a preferred tool to measure the axial and lateral velocity components through bundles due to its accuracy. Although LDV is a mature technique, it only provides point measurements which do not allow for the instantaneous investigation of the momentum exchange among neighboring subchannels; moreover, the small spaces between the rods and within the subchannels makes LDV measurements difficult. In the other hand, flow visualization techniques can provide a 2D map of the instantaneous and average structure of the liquid velocity. Among the used visualization techniques, PIV has gained a lot of attention due to its high accuracy, relative simple application, and non-intrusiveness. Several studies had applied PIV to study the flow structures within bundles sub-channels with some success; however, in these studies there exist an increased uncertainty due to reflections and shadows from the rods. Shadows and undesired reflections appear when the light passes through materials with different refractive indexes. This issue was addressed in other studies on which the refractive index of the rod bundle (or sections of it) was matched with that of water. These studies used particle tracking velocimetry (PTV) and obtained comparable results with LDV measurements. This success indicated that RIM-PTV experiments can be successfully implemented to gather information for simulation benchmarking. The present work is an extension of previous efforts on which measurements were performed only on vertical planes but lacked of the important flow structures that can only be observed through a horizontal cross section of the bundle at different axial height. The focus of the present study was to fill this information gap by measuring the lateral velocity in horizontal planes at different heights. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Neutron-Gamma flux and Dose calculations in a Pressurized Water Reactor (PWR)

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to the French nuclear safety authorities. In this context, one of its missions is to improve nuclear safety through the enhancement of knowledge regarding operating nuclear power reactors. The issues concerning the aging and the characterisation of the reactor vessel require knowledge of the neutron and gamma radiation environment on the vessel and beyond. Since few years, ENEA (Energia Nucleare e Energie Alternative, i.e. Nuclear Energy and Alternative Energy agency) and IRSN collaborate on Monte-Carlo simulations and Variance Reduction Techniques (VRT), which are used to determine neutron and gamma spectra in and out of the core. Additionally, IRSN participates in the DISCOMS project, which stands for 'Distributed Sensing for Corium Monitoring and Safety', aiming at examining the status of the third barrier of confinement, in order to define possible mitigation strategies in case of a severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) reactor pit concrete erosion and iii) corium cooling. One of the IRSN tasks in the frame of the DISCOMS project is to provide the radiation level in terms of neutron and gamma doses and flux energy spectra at various locations. This paper summarizes the study performed on the radiation transport in the Tihange-I Belgian nuclear reactor and contributes to several issues requiring the knowledge of the radiation environment in and outside the reactor vessel. This paper gives an overview of the work presented at the ICRS-13 and RPSD-2016 conference (for more details see the proceedings paper)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

RISMC Industry Application no. 1 (ECCS/LOCA) Core characterization automation: Reference PWR designs for IA no. 1

The U.S. NRC is currently proposing rulemaking '10 CFR 50.46c' to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that both cladding and ECCS performance need to be considered in a coupled manner. This implies that LOCA methodologies will need to be updated for regulatory purposes. The Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) research and development program initialed several programs at the Idaho National Laboratory (INL) to support the industry by offering potential solutions to this new LOCA/ECCS analysis. One of these programs is the Industry Application no. 1 (IA1). It's goal is to develop an integrated evaluation model called LOCA Toolkit for the US (LOTUS) to revisit how risks are evaluated in a LOCA under 10 CFR 50.46c. The focus of LOTUS is to establish the automation interfaces among the physics disciplines involved, namely the core design automation, the fuels/clad performance, the system analysis, the risk assessment and the core design optimization. The risk-informed margins management approach for IA1 will provide a means of quantifying the impact on the key LOCA analysis figures-of-merit like peak cladding temperature (PCT), equivalent cladding reacted (ECR), etc. of a change in LOCA analysis inputs. The information obtained can then be used for decision-making and margin management. The modular approach for LOTUS will enable plant owners/vendors to further customize the LOTUS framework for use within their established codes and methods. One of the LOTUS modules deals with the core design automation (CD-A). The purpose of this module is to supply the LOCA analysis with initial conditions, i.e. assembly/pin power histories, power shapes, etc. to be employed by the Fuels/Clad Performance (EP) and System Analysis (SA) modules. In the CD-A module, the plant owner/operator will characterize his core design with the tools he has. For the current LOTUS demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D2,3. Cross section generation will be done using the Studsvik lattice code HELIOS-2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A Fully Passive Daily Load-follow Operation in a Small PWR System

An innovative fully passive load-follow operation (PLFO) concept, based only on the negative reactivity feedbacks, is proposed by the authors. The new design concept requires no control rod adjustment or boron utilization for achieving load-follow operation in pressurized water reactors (PWRs). PLFO is motivated by the need for developing next-generation PWRs with enhanced power maneuvering capabilities, increased level of autonomism, and less dependence on active controllers. Enhanced power maneuvering capabilities are required to allow next-generation PWRs to actively contribute to electrical grid stability. Meanwhile, increased level of autonomism and less dependence on active controllers are required for safety and reliability of next-generation PWRs. However, developing an autonomous thermal reactor is challenging due to its strong thermal and xenon (Xe) reactivity feedback. Therefore, enhancing power maneuvering capabilities and simultaneously increasing the level of autonomy with less dependence on active controllers is very challenging in PWRs. However, PLFO is a promising solution. For the feasibility of PLFO, the coolant temperature coefficient (CTC) should be always clearly negative and this requires a soluble boron-free (SBF) coolant system. An SBF coolant system is also favorable in terms of reactor safety. Nevertheless, utilizing soluble boron in load-following operation is not favorable, as the change in boron concentration is a quite slow process. The power demand can be either regular (repeated pattern in a daily-basis) or irregular (varying demand). The feasibility of PLFO for regular demand was demonstrated by the authors in other works. This paper is dedicated for summarizing the feasibility of PLFO for irregular demand, and to point out the impact of Xe reactivity on PLFO for a varying power demand.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Effect of nickel on the oxidation behavior of FeCrAl alloy in simulated PWR and BWR conditions

FeCrAl alloys are being investigated as candidate materials for light water reactor fuel claddings. The chemical composition of FeCrAl determines its resistance to corrosion under normal operating conditions. The purpose of this paper is to investigate the effect of 0, 1, 3 wt% Ni addition on the passivation characteristics of Fe17Cr5.5Al in oxygenated and hydrogenated simulated reactor waters for 3-month immersion. Current results show that Ni addition improves the passivation of FeCrAl under oxygenated condition and decreases the mass loss under hydrogenated conditions.

36 MATERIALS SCIENCE↗