High Flux Isotope Reactor Accident Analysis
High Flux Isotope Reactor Accident Analysis

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Published: 1977

Total Pages:

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In 1968 the ion chambers in the HFIR safety system were repositioned (moved further from the core) in order to extend chamber lifetime and increase system reliability. In 1973 new flux signal conditioners were installed to replace obsolete components. As a result of these two modifications, the system has become more sensitive to noise. Efforts to eliminate the noise at the source met with little or no success. Remedial changes are described and evaluated. Specifically, these changes include placing feedback smoothing capacitors in the ion chamber flux amplifiers and removing the rate trip in Modes 2 and 3. It is the purpose of the supplement to update the original safety analysis document, to describe the changes, and to provide a summary of the performed analyses which show that the safety implications of the changes are acceptable. These changes fall within the scope of the original safety analysis document and the supplement substantiates that the changes described do not involve a safety question not reviewed in the original safety analysis document.

Large Break Loss-of-coolant Accident Analyses for the High Flux Isotope Reactor
Large Break Loss-of-coolant Accident Analyses for the High Flux Isotope Reactor

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Published: 1989

Total Pages: 10

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The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

Large Break Loss of Coolant Severe Accident Sequences at the HFIR (High Flux Isotope Reactor).
Large Break Loss of Coolant Severe Accident Sequences at the HFIR (High Flux Isotope Reactor).

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Published: 1990

Total Pages: 30

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An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

A System Analysis Computer Model for the High Flux Isotope Reactor (HFIRSYS Version 1).
A System Analysis Computer Model for the High Flux Isotope Reactor (HFIRSYS Version 1).

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Published: 1992

Total Pages: 128

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A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear's stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system's behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

Fracture Capacity of High Flux Isotope Reactor (HFIR) Vessel with Random Crack Size and Toughness
Fracture Capacity of High Flux Isotope Reactor (HFIR) Vessel with Random Crack Size and Toughness

Author: S-J Chang

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Published: 1994

Total Pages: 16

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The probability of fracture versus a range of applied hoop stresses along the high flux isotope reactor (HFIR) vessel is obtained as an estimate of its fracture capacity. Both the crack size and the fracture toughness are assumed to be random variables and subject to assumed distribution functions. Possible hoop stress is based on the numerical solution of the vessel response by applying a point pressure-pulse at the center of the fluid volume within the vessel. Both the fluid-structure interaction and radiation embrittlement are taken into consideration. Elastic fracture mechanics is used throughout the analysis. The probability function of fracture for a single crack due to either a variable crack depth or a variable toughness is derived. Both the variable crack size and the variable toughness are assumed to follow known distributions. The probability of vessel fracture with a multiple number of cracks is then obtained as a function of the applied hoop stress. The probability of fracture function is, then, extended to include different levels of confidence and variability. It, therefore, enables one to estimate the high confidence and low probability fracture capacity of the reactor vessel under a range of accident loading conditions.

Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor
Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

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Published: 2015

Total Pages: 46

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The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.