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Tuesday, August 4, 2020 | History

2 edition of Modelling of the thermal and mechanical performance of a fast reactor control rod pin. found in the catalog.

Modelling of the thermal and mechanical performance of a fast reactor control rod pin.

Graham James Duffy

Modelling of the thermal and mechanical performance of a fast reactor control rod pin.

by Graham James Duffy

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Published by Universityof Salford in Salford .
Written in English


Edition Notes

MSc thesis, Physics.

ID Numbers
Open LibraryOL20905803M

zoning in fuel, blanket & control rod region is shown in Fig. The power distribution and flow for a 30o core sector are shown in Fig Temperature rise across the hottest channel and SA are shown in Fig The clad midwall hotspot and nominal temperatures are shown in Fig The values are for a typical fast reactor. Relap5 is used to model the SPWR. Steady-state simulation is performed to verify the correctness of the model. In the reactor power control system, both the reactor fission power and the coolant average temperature are regulated by the control rod .

  thermal-mechanical nuclear fuel performance code to support licensing reviews for non-light water reactor (LWR) fuel designs, and the code development activities needed to adequately capture that physics. Unlike the other volumes in the NRC Non-Light Water Reactor . THERMAL-HYDRAULIC ANALYSIS OF A 7-PIN SODIUM FAST REACTOR FUEL BUNDLE WITH A NEW PATTERN OF HELICAL WIRE WRAP SPACER Seong Dae Park, Sung Bo Moon, Seok Bin Seo, In Cheol Bang* School of Mechanical and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST) 50 UNIST-gil, Ulju-gun, Ulsan, , Republic .

Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and. of the cooling system are located on the roof of the reactor building. A bird’s eye view of the HTTR facility is given in Fig The reactor core is designed to generate 30 MW of thermal power and consists of an array of hexagonal graphite fuel assemblies so-called “pin-in-block type fuel”(See Fig. 4), control rods and graphite reflector.


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Modelling of the thermal and mechanical performance of a fast reactor control rod pin by Graham James Duffy Download PDF EPUB FB2

The current modelling of thermal properties of fast reactor MOX in fuel performance codes (FPCs, e.g., GERMINAL, TRANSURANUS, BISON, FEMAXI) is limited, for various reasons. First, available experimental data regard fresh MOX or LWR irradiation conditions, while only very few data correspond to irradiated fast reactor : A.

Magni, T. Barani, A. Del Nevo, D. Pizzocri, D. Staicu, P. Van Uffelen, L. Luzzi. The characteristics of Generation-IV fast reactors strongly impact the design of its control rods.

The traditional control rod is a cluster of vented pins with boron carbide (B 4 C) as the absorber. Due to the gas release, the neutron-induced swelling, the melting risk, and the high loss of the reactivity worth, B 4 C impact the safety performance of control rods and thus Cited by: 2.

Sodium-cooled fast reactor pin model for predicting pin failure during a power excursion the fast increase of nuclear power induces a strong temperature rise in the fuel pellets leading to strong mechanical and thermal loads on the cladding which could lead to clad failure or/and fuel meltdown.

such as control rod withdrawal, and fast Cited by: 5. Incorporating global flux and depletion solvers, subchannel thermal-hydraulics codes, pin-level power and flux reconstruction methods, detailed fuel cycle and history tracking systems, finite element-based fuel performance coupling, reactivity coefficient generation, SASSYS-1/SAS4A transient modeling, control rod worth routines, and multi Cited by: 1.

Fast reactor design requires the simultaneous application of mechanical and thermal-hydraulics analysis methods. We reviewed some aspects of mechanical analysis in Chapter 8; we will discuss mechanical design further in Chapter Chapters 9 and 10 will deal with thermal-hydraulics : Alan Waltar, Donald Todd.

In the UK fast reactor irradiation programme, a Nimonic alloy has shown good performance both as fuel pin cladding and as wrapper tube material. Fuel pins, clad with this Nimonic alloy, are achieving full burnup in excess of 15% heavy atom during irradiation in the Prototype Fast Reactor (PFR) at Dounreay.

cooled Fast Reactors (SFR). This core thermal-hydraulic analysis is performed at three scales: Individual sub-assemblies, characterized by its triangular-lattice pin bundle with helical spacer wires: In the s, a specific subchannel scale model for SFR-type subassemblies was developed at CEA.

rod bundle thermal hydraulic analysis of the nuclear reactor core as well as heat exchangers, namely, (a)-subchannel analysis, (b)-porosity and distributed resistance approach and finally (c)-benchmark rod-bundle analysis which uses a boundary fitted coordinate system.

The first approach is widely used in the subchannel codes. To prevent a metallic fuel rod failure in a fast reactor, it is required to evaluate the design limits such as (1) cladding strain and cumulative damage fraction (CDF), (2) fuel melting, and (3) eutectic melting.

The design requirement for cladding is assumed to be 1% of the thermal creep strain and of CDF. Thermal-mechanical modelling. Dimensions of the core vary due to temperature change of structural elements.

The expansion mechanism is not always straightforward and depends on the actual state of the core. In this paper, we consider only the fresh core of the ALLEGRO reactor, which simplifies the thermal–mechanical modelling.

3. Behaviour under irradiation Gas production and -release, and swelling from the absorber material are the most important irradiation effects for control rod performance. Moreover, there is a known influence of irradiation on thermal conductivity of B4C, which is described in chapter (thermal)–1MW (electric) U–Pu–Zr metal fueled, sodium cooled, fast spectrum reactor with lithium inlet and outlet temperatures of and C, respectively.

The nuclear power system performance parameters of RAPID are shown in Table 1. This power level is sufficient for private residences. The reactor structure is shown in Fig.

The gamma-ray dose rate has to be measured in dependence on thermal reactor power at selected test points in the reactor hall. Critical control rod positions have to be determined and discussed in dependence on the thermal power of the reactor. Types of Reactor Start-up Starting-up a nuclear reactor means to generate a controlled.

Control rod Example of the plant 4. Plant stability Channel hydraulic stability and core stability Ledinegg instability Density wave oscillation Geysering Chugging 5. Application of component modeling to the nuclear power plant Plant transient in Liquid-metal-cooled fast reactors Heat transfer between.

• Rod Cluster Control: RCC • Reactor Internals: RI • Reactor Vessel: RV • Control Rod Drive Mechanism: CRDM (1) FUEL ASSEMBL Y The reactor core consists of a specified number of fuel rods that are held in bundles by spacer grids and top and bottom nozzles.

Fuel assemblies are arranged in predetermined square matrix patterns Figure   The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss.

reactor to remove the left out arrive at the adequate design of the reactor, temperature distribution needs to be accurately predicted inside the reactor.

This paper describes the thermal modeling of chlorination reactor using COMSOL Multiphysics. impurities and the un-reacted chlorine gas is. PWR MOX Rod Ejection Accident (REA) benchmark is utilized in the presented study. KEYWORDS Burnup and Gadolinium content, fuel-thermal-conductivity model, dynamic-gap- conductance model, feedback effects.

INTRODUCTION In this paper, the study of the Rod Ejection Accident (REA), a design basis reactivity insertion accident. This would require core-level modeling of coupled thermal-hydraulics and spatial kinetics. The characteristics of a fast-acting shutdown system widely deployed in CANDU reactors (in particular, shut down system 1—rod insertion) would serve as the basis for a similar system in the PT-SCWR for the purpose of this study.

PCI is controlled by the complex interplay of thermal, mechanical, and chemical behavior of a fuel rod during plant operation; thus modeling PCI requires an integral fuel performance. Sodium-cooled Fast Reactor used for minor actinides transmutation neither the model for the control rod expansion feedback nor the model for negligible manner the fuel thermal-mechanical.A 2D Transient Model for Gas-Cooled Fast Reactor Plate-Type Fuel P.

Petkevich *1, 2, K. Mikityuk1, P Abstract – An accurate analysis of the thermal-mechanical behaviour of the fuel is particularly important for advanced reactor systems due to their increased safety requirements. (system thermal hydraulics) and FRED [6] (fuel pin.Research reactor control rods are composed of materials which strongly absorb and so, without special methods, diffusion theory cannot be used to calculate control rod worths in thermal research reactors.

Higher order methods, such as Monte Carlo techniques, are commonly used for blackness coefficients for the fast groups are not needed.