Cyclic Operation of Pressure Piping With γ Heating

1960 ◽  
Vol 82 (2) ◽  
pp. 447-451 ◽  
Author(s):  
K. R. Merckx
Keyword(s):  
Heat Generation ◽  
Reactor Core ◽  
Operating Conditions ◽  
Pressure Tube ◽  
Maximum Pressure ◽  
Residual Tensile Stress ◽  
Tube Surface ◽  
Elastic Plastic ◽  
Cyclic Operation ◽  
Test Reactor

An elastic-plastic analysis is developed for an internally cooled pressure tube with uniform heat generation. This analysis extends the method of calculating the location of the elastic-plastic boundary reported by Barrie [4] to account for the change in the plastic zone due to residual stresses which occur during cyclic operation. Numerical calculations are made for operating conditions expected to be encountered in a pressure tube in a loop through the Engineering Test Reactor Core. The numerical results show that the radius of the initial plastic boundary decreases during subsequent loading cycles. Also, for equal maximum pressure and volumetric heat generation, the total plastic strain per operational cycle on the inner tube surface and the residual tensile stress on the outer tube surface increase when the tube wall is thickened.

Assessing Advanced Test Reactor Process States Using Acoustic Measurement Infrastructure

2017 ◽  
Author(s):  
Vivek Agarwal ◽  
James A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Signal Transmission ◽  
National Laboratory ◽  
Reactor Core ◽  
Operating Conditions ◽  
Acoustic Measurement ◽  
Primary Coolant ◽  
Test Reactor ◽  
Operation Cycle ◽  
Measurement Infrastructure

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement). This paper demonstrated different in acoustic signature of the ATR under different operating conditions. In particular, ATR acoustic baseline is captured during typical operation cycle and during power axial locator mechanism operation cycle. The difference in two acoustic baseline is significant and highlights salient difference that are critical in the design and development of acoustically telemetered sensors.

Finite element method estimation of temperature distribution of automotive tire using one-way-coupled method

2021 ◽  
pp. 095440702110087
Author(s):  
Zumrat Usmanova ◽  
Emin Sunbuloglu
Keyword(s):  
Thermal Analysis ◽  
Temperature Distribution ◽  
Heat Generation ◽  
Complex Geometry ◽  
Rolling Resistance ◽  
Operating Conditions ◽  
Deformation Analysis ◽  
Coupled Method ◽  
Experimental Values ◽  
Hysteretic Loss

Numerical simulation of automotive tires is still a challenging problem due to their complex geometry and structures, as well as the non-uniform loading and operating conditions. Hysteretic loss and rolling resistance are the most crucial features of tire design for engineers. A decoupled numerical model was proposed to predict hysteretic loss and temperature distribution in a tire, however temperature dependent material properties being utilized only during the heat generation analysis stage. Cyclic change of strain energy values was extracted from 3-D deformation analysis, which was further used in a thermal analysis as input to predict temperature distribution and thermal heat generation due to hysteretic loss. This method was compared with the decoupled model where temperature dependence was ignored in both deformation and thermal analysis stages. Deformation analysis results were compared with experimental data available. The proposed method of numerical modeling was quite accurate and results were found to be close to the actual tire behavior. It was shown that one-way-coupled method provides rolling resistance and peak temperature values that are in agreement with experimental values as well.

Derivation of Input Distributions for Probabilistic Evaluation Using In-Reactor Measurements

2012 ◽  
Author(s):  
Eric Nadeau
Keyword(s):  
Risk Reduction ◽  
Management Strategy ◽  
Reactor Core ◽  
Pressure Tube ◽  
Channel Management ◽  
Maintenance Strategy ◽  
Acceptance Criteria ◽  
Fuel Channel ◽  
Probabilistic Evaluation ◽  
Commercial Operation

Candu Energy Inc. (former commercial operation of AECL) has developed probabilistic tools to support nuclear plant operators with a risk-based fuel channel management strategy. One such tool is used to evaluate the probability of pressure tube rupture resulting from pressure tube to calandria tube contact and hydride blisters. This tool assumes that PT rupture occurs when delayed hydride cracking (DHC) initiates in a blister. The objectives of the probabilistic assessments are to: • Determine the overall risk of PT rupture in the reactor core for comparison with the acceptance criteria. • Determine the risk of PT rupture for specific fuel channels to assist in the development of an inspection/maintenance strategy. • Evaluate the risk reduction that would result from different fuel channels inspection/maintenance scenarios. • Optimize inspection/maintenance programs. The distributions of the most critical input distributions can be derived by benchmarking against in-reactor measurements. Two benchmark methods were developed to take advantage of the recent advancements in the accuracy of the inspection tool that measures the gap profile between the PT and the CT.

Experiences in Examining and Dispositioning Flaws in Zr-2.5Nb Pressure Retaining CANDU Reactor Components

2002 ◽  
Author(s):  
Andrew Celovsky ◽  
John Slade
Keyword(s):  
Service Life ◽  
Reactor Core ◽  
Pressure Tube ◽  
Candu Reactors ◽  
Periodic Inspection ◽  
Pressure Tubes ◽  
Surveillance Programs ◽  
Non Destructive ◽  
Limited Service ◽  
Non Destructive Techniques

CANDU reactors use Zr-2.5 Nb alloy pressure tubes, as the primary pressure boundary within the reactor core. These components are subject to periodic inspection and material surveillance programs. Occasionally, the inspection program uncovers a flaw, whereupon the flaw is assessed as to whether it compromises the integrity of the pressure-retaining component. In 1998, such a flaw was observed in one pressure tube of a reactor. Non-destructive techniques and analysis were used to form a basis to disposition the flaw, and the component was fit for a limited service life. This component was eventually removed from service, whereupon the destructive examinations were used to validate the disposition assumptions used. Such a process of validation provides credibility to the disposition process. This paper reviews the original flaw and its subsequent destructive evaluation.

Seismic Study of High-Temperature Engineering Test Reactor Core Graphite Structures

1992 ◽  
Vol 99 (2) ◽  
pp. 158-168 ◽  
Author(s):  
Tatsuo Iyoku ◽  
Yoshiyuki Inagaki ◽  
Shusaku Shiozawa ◽  
Isoharu Nishiguchi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Test Reactor

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

SIMULASI KARAKTERISTIK ALIRAN DAN SUHU FLUIDA PENDINGIN (H2O) PADA TERAS REAKTOR NUKLIR SMR (SMALL MODULAR REACTOR)

ROTASI ◽  
2013 ◽  
Vol 15 (4) ◽  
pp. 33
Author(s):  
Anwar Ilmar Ramadhan ◽  
Indra Setiawan ◽  
M. Ivan Satryo
Keyword(s):  
Fluid Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Axial Direction ◽  
Operating Conditions ◽  
Fluid Temperature ◽  
Wide Range ◽  
Fuel Reactor ◽  
Performance Of Heat Transfer

Safety is an issue that is of considerable concern in the design, operation and development of a nuclear reactor. Therefore, the method of analysis used in all these activities should be thorough and reliable so as to predict a wide range of operating conditions of the reactor, both under normal operating conditions and in the event of an accident. Performance of heat transfer to the cooling of nuclear fuel, reactor safety is key. Poor heat removal performance would threaten the integrity of the fuel cladding which could further impact on the release of radioactive substances into the environment in an uncontrolled manner to endanger the safety of the reactor workers, the general public, and the environment. This study has the objective is to know is profile contour of fluid flow and the temperature distribution pattern of the cooling fluid is water (H2O) in convection in to SMR reactor with fuel sub reed arrangement of hexagonal in forced convection. In this study will be conducted simulations on the SMR reactor core used sub channel hexagonal using CFD (Computational Fluid Dynamics) code. And the results of this simulation look more upward (vector of fluid flow) fluid temperature will be warm because the heat moves from the wall to the fluid heater. Axial direction and also looks more fluid away from the heating element temperature will be lower.

scholarly journals Closed-loop control of a dual-fuel engine working with different combustion modes using in-cylinder pressure feedback

2019 ◽  
Vol 21 (3) ◽  
pp. 484-496 ◽  
Author(s):  
Carlos Guardiola ◽  
Benjamín Pla ◽  
Pau Bares ◽  
Alvin Barbier
Keyword(s):  
Closed Loop ◽  
Combustion Engine ◽  
Fuel Combustion ◽  
Operating Conditions ◽  
Dual Fuel ◽  
Maximum Pressure ◽  
Cylinder Pressure ◽  
Pressure Derivative ◽  
Combustion Modes ◽  
Dual Fuel Combustion

This work presents a closed-loop combustion control concept using in-cylinder pressure as a feedback in a dual-fuel combustion engine. At low load, reactivity controlled compression ignition combustion was used while a diffusive dual-fuel combustion was performed at higher loads. The aim of the presented controller is to maintain the indicated mean effective pressure and the combustion phasing at a target value, and to keep the maximum pressure derivative under a limit to avoid engine damage in all the combustion modes by cyclically adapting the injection settings. Various tests were performed at steady-state conditions showing good abilities to fulfil the expected operating conditions but also to reject disturbances such as intake pressure or exhaust gas recirculation variations. Finally, the proposed control strategy was tested during a load transient resulting in a combustion switching-mode and the results exhibited the closed-loop potential for controlling such combustion concept.

An Experimental Analysis of Misalignment Effects on Hydrodynamic Plain Journal Bearing Performances

2001 ◽  
Vol 124 (2) ◽  
pp. 313-319 ◽  
Author(s):  
J. Bouyer ◽  
M. Fillon
Keyword(s):  
Film Thickness ◽  
Journal Bearing ◽  
Hydrodynamic Pressure ◽  
Operating Conditions ◽  
Temperature Fields ◽  
Maximum Pressure ◽  
Minimum Film Thickness ◽  
Oil Flow ◽  
Hydrodynamic Effects

The present study deals with the experimental determination of the performance of a 100 mm diameter plain journal bearing submitted to a misalignment torque. Hydrodynamic pressure and temperature fields in the mid-plane of the bearing, temperatures in two axial directions, oil flow rate, and minimum film thickness, were all measured for various operating conditions and misalignment torques. Tests were carried out for rotational speeds ranging from 1500 to 4000 rpm with a maximum static load of 9000 N and a misalignment torque varying from 0 to 70 N.m. The bearing performances were greatly affected by the misalignment. The maximum pressure in the mid-plane decreased by 20 percent for the largest misalignment torque while the minimum film thickness was reduced by 80 percent. The misalignment caused more significant changes in bearing performance when the rotational speed or load was low. The hydrodynamic effects were then relatively small and the bearing offered less resistance to the misalignment.

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