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Sunday, August 2, 2020 | History

3 edition of Analysis of pin-by-pin effects for LWR rod ejection accident found in the catalog.

Analysis of pin-by-pin effects for LWR rod ejection accident

Analysis of pin-by-pin effects for LWR rod ejection accident

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Published by Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Supt. of Docs., U.S. G.P.O. [distributor in Washington, DC .
Written in English

    Subjects:
  • Light water reactors -- Accidents -- Computer simulation.,
  • Nuclear fuel rods -- Accidents -- Computer simulation.

  • Edition Notes

    Other titlesAnalysis of pin by pin effects for LWR rod ejection accident.
    Statementprepared by A. Avvakumov, V. Maloveev, V. Sidorov.
    SeriesInternational agreement report -- NUREG/IA-0175.
    ContributionsMalofeev, V., Sidorov, V., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research., Institut problem bezopasnogo ispolʹzovanii͡a i͡adernoĭ ėnergii (Rossiĭskiĭ nauchnyĭ t͡sentr "Kurchatovskiĭ institut")
    The Physical Object
    FormatMicroform
    Paginationxi, 77 p.
    Number of Pages77
    ID Numbers
    Open LibraryOL15564521M

    The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3 x 3 assembly mini-core and a full pressurized water reactor (PWR) core. The analysis of the behavior of light water reactor (LWR) fuel rods is described. The properties of relevant fuel and cladding materials are discussed and numerical data are given. The basic phenomena taking place in pellet-in-cladding nuclear reactor fuel are described systematically, including neutronic aspect of the fuel, the thermal and.

    "This study was undertaken to demonstrate capabilities of the pin-by-pin model used by the BARS code and to understand various effects of intra-assembly pin-by-pin representation of fuel power, bumup and temperature in calculational analysis of light water reactor rod ejection accidents (LWR REAs). Effects of pin-by-pin fuel power and bumup. |a Main stream line break analysis for lungmen ABWR / |c prepared by Chunkuan Shih [and five others]. 1 |a Washington, DC: |b Division of Systems Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, |c June |a 1 online resource (xiii, 24 pages): |b illustrations + |e errata.

    Reactor Vessel Closure Head Drop Analysis: Sensitivity Study on the Effects of Representing Nonlinear Behavior in the Closure Head Assembly. Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Technique. Tewfik Hamidouche, Break-Up of Gas Stratification in LWR Containment Induced by Negatively Buoyant Jets and Plumes. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP 3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3 x 3 assembly mini-core and a full pressurized water reactor (PWR) by: 3.


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Analysis of pin-by-pin effects for LWR rod ejection accident Download PDF EPUB FB2

Effects of pin-by-pin fuel power and bumup representation were investigated on the basis of calculations for the peripheral control rod ejection in VVER of the South Ukrainian NPP Unit 1. Comparative analysis of the REA in pressurized water reactor (PWR) of Three Mile Island Unit 1 using the BARS code with the diffusion nodal codes PARCS and.

by the BARS code and to understand various effects of intra-assembly pin-by-pin representation of fuel power, bumup and temperature in calculational analysis of light water reactor rod ejection accidents (LWR REAs).

Avvakumov, V. Malofeev, and V. Sidorov, “Analysis of Pin-by-Pin Effects for LWR Rod Ejection Accident,” NUREG/IA, NSI RRC KI /, IPSN/, 2.

Avvakumov, V. Malofeev, and V. Sidorov, “Uncertainty Analysis for PWR Rod Ejection Accident Using the RELAP-BARS Code,” Nuclear Safety Institute of Russian. Analysis of pin-by-pin effects for LWR rod ejection accident book this from a library.

Analysis of pin-by-pin effects for LWR rod ejection accident. [A Avvakumov; V Malofeev; V Sidorov; U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research.; Institut problem bezopasnogo ispolʹzovanii︠a︡ i︠a︡dernoĭ ėnergii (Rossiĭskiĭ nauchnyĭ t︠s︡entr.

This study was undertaken to demonstrate capabilities of the pin-by-pin model used by the BARS code and to understand various effects of intra-assembly pin-by-pin representation of fuel power, bumup and temperature in calculational analysis of light water reactor rod ejection accidents (LWR REAs).

Cofrentes NPP (BWR/6) ATWS (MSIVC) Analysis with TRAC-BF1: Mar ML NUREG/IA Analysis of Pin-by-Pin Effects for LWR Rod Ejection Accident: Mar ML NUREG Standard Review Plan for Spent Fuel Dry Storage Facilities: Mar ML NUREG Reactivity-initiated accident (RIA) is a postulated accident that involves an undesirable increase in fission rate and reactor power.

The power increase may damage the reactor core, and in severe cases, even may lead to disruption of the reactor. The design basis reactivity accident in PWR is the rod ejection by: 3. The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel.

The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. The following example refers to a generic rod ejection accident (REA) analysis of a Siemens built PWR with MW thermal power and fuel assemblies of the type 18 × 18– A rod ejection accident is simulated by means of PANBOX at the end of a representative cycle with an initial power of 30% P / P N (P / P N = rated nominal power).Cited by: 3.

Analysis of pin-by-pin effects for LWR rod ejection accident / by: Avvakumov, A., et al. Published: () TRACE (V Patch 2) validation based on the RELAP5-Calculation of FIX-III LOCA experiments no./ by: Sheng, S. ChunHong, Published: (). This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident.

Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological by: 7. The rod ejection accident (REA) is the design basis accident in pressurize water reactor. It is as-sumed a mechanical failure of the control rod drive mechanism housing that causes control rod as-sembly ejection from the reactor core in approximately s.

Analyses of the accident is necessary for getting license to operate nuclear power plant. Analysis of pin-by-pin effects for LWR rod ejection accident / by: Avvakumov, A., et al.

Published: () Thermal hydraulic and fuel rod mechanical combination analysis of Kuosheng Nuclear Power Plant with RELAP5 MOD/FRAPTRAN/Python in SNAP interface / by: Wang, Jong-Rong, Published: ().

In this paper we present a multi-physics neutronics-fuel thermal-thermal hydraulics uncertainty analysis methodology for Rod Ejection Accident (REA) in. A Super LWR with double tube water rods has been designed for simplifying the upper core structure (Wu and Oka, ).The coolant flow in double tube water rod is shown in Fig.

1(b). The water flows upward in the inner water rod and then flows downward through the outer water by: 3. Design-basis accidents such as control rod ejection accident (REA) must be evaluated as part of a deterministic safety analysis for any reactor licensing proce. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP 3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem.

The analysis is performed on two numerical benchmarks, a 3 × 3 assembly mini-core and a full pressurized water reactor (PWR) by: 3. Request PDF | Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Technique | This paper investigates the possibility to extend standard computer tools and methods, commonly.

Analysis of Rod Ejection Accident in a Research Reactor by the Coupled Code Technique Conference Paper (PDF Available) March with 1, Reads How we measure 'reads'. A three-dimensional pin-by-pin nodal-transport code for a pressurized water reactor (PWR) core analysis, SCOPE2, was used in this study since it can directly treat the pin-by-pin feedback effect.

plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor.

Both power and temperature pulse following the reactivity- initiated accidents are calculated.Analysis of Reactivity - Initiated Accident for Control Rods Ejection delayed precursor groups, resulting in a syst em consisting of seven coupled differential equations.

Obtaining accurate results is often problematic, because the equations are stiff with many techniques, where very small time steps are used.Search for books, ebooks, and physical FRAPTRAN a computer code for the transient analysis of oxide fuel rods / Bibliographic Details; Corporate Authors: U.S.

Nuclear Regulatory Commission. Analysis of pin-by-pin effects for LWR rod ejection accident / .