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RIA 1-4 Experiment Specification Document

Description: The Reactivity Initiated Accident (RIA) test series is being performed in the Power Burst Facility (PBF) to provide data for verifying analytical codes capable of predicting light water reactor fuel performance during hypothetical control rod drop (or ejection) accidents. The most important aspect of an RIA or an RIA test is the magnitude of energy deposited into the fuel, therefore the primary purpose of the RIA test series is to better define the relationship between e.nergy deposition and fuel rod behavior. In particular, the RIA research objectives are to determine failure thresholds, modes, and consequences with respect to total energy deposition, irradiation history, and fuel design. The most severe RIA is the postulated Boiling Water Reactor (BWR) control rod drop during reactor startup, therefore all the RIA tests will be conducted at BWR startup coolant conditions. Test RIA 1-4 will consist of a 4 x·4 array of rods positioned in a coolant flow shroud. Fourteen of the rods are unirradiated BWR/6-type fuel rods and two are water-filled rods. Two water rods will be interior rods positioned along one diagonal. The primary objective of Test RIA 1-4 will be to obtain data on clustered fuel rod behavior during a rapid power transient simulating a BWR control rod drop. This document provides a basis for understanding the test plan for Test RIA 1-4. Past RIA test experience and a review of the PBF-RIA tests which are scheduled for completion prior to Test RIA 1-4 are summarized. The fuel rod, water rod, grid spacer, flow channel, and bundle support structure specifications are described. The instrumentation required for this test is discussed. The preliminary reactor operation and posttest operation requirements are presented.
Date: September 1, 1978
Creator: Semken, R. S.
Partner: UNT Libraries Government Documents Department

LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

Description: Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.
Date: February 1, 1981
Creator: Bennett, W. D.; Goodman, R. L.; Heaberlin, S. W.; Hesson, G. M.; Nealley, C.; Kirg, L. L. et al.
Partner: UNT Libraries Government Documents Department

Reactivity Initiated Accident Test Series RIA Scoping Test Experiment Operating Specification

Description: This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Scoping Test to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The primary objectives of the RIA research are to determine fuel failure thresholds, modes, and consequences as functions of (a) enthalpy insertion, (b) irradiation history, and (c) fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors (BWRs) will be used in the first six RIA tests, termed Series I.
Date: June 1, 1978
Partner: UNT Libraries Government Documents Department

Irradiated Tube Testing Facility

Description: A logical sequence for evaluating the mechanical properties of N-Reactor pressure tubing in the as-received condition is similar to evaluation after reactor service. The usual difficulties associated with testing a heavy tube to failure with 300 C water are compounded after reactor service by the high gamma radioactivity of the discharged tubes. Success in testing of discharged reactor pressure tubing in a pilot underwater facility led to the construction of an Irradiated Tube Testing ·Facility (ITTF) adjacent to the Radiometallurgy Building. The ITTF consists of a 10 .ft 6 in. deep, water-filled basin 15 ft long by 10 ft wide, with six test chambers placed within the basin. These chambers are stainless steel bell jars that enclose the specimen, blast shielding, and electric furnace. The vessels exclude water from the hot specimen and furnace during testing and contain the pressure surge that occurs when the hot pressurized water flashes to steam upon failure of the test specimen. The ITTF basin is connected to the Radiometallurgy Building Basin by an underwater tube containing a conveyor system. The conveyor makes it possible to transfer highly radioactive materiais with ease and safety. Remote handling tools were developed to measure the tubes, attach end closures, attach thermocouples, and load lhe specimen within the furnace containment vessels.
Date: November 1, 1964
Creator: Jackson, P. M.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The schedule for the installation of the PAR slurry loop experiment in the ·South Facility af the ORR has been reviewed and revised. Changed design philosophy to include maintainability of the loop auxiliaries required extension of the main construction period to the end of,June, with a three month period ending in September, for test and run in during which construction will be completed.
Date: February 28, 1958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY PERIODIC TECHNICAL REPORT MARCH, APRIL, MAY, JUNE 1970

Description: This report was prepared by Battelle-Northwest under Contract No. AT (45-1) -1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during March, April , May and June 1970.
Date: August 1, 1970
Creator: Wolfe, B. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department

Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

Description: This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.
Date: September 13, 2013
Creator: Tokuhiro, Akira & Jones, Byron
Partner: UNT Libraries Government Documents Department

Reactivity Initiated Accident Test Series Test RIA 1-1 Experiment Operating Specification

Description: This document describes the experiment operating specifications for the Reactivity Initiated Accident (RIA) Test RIA 1-1 to be conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The RIA Series I research objectives are to determine fuel failure thresholds, modes and consequences as functions of enthalpy insertion, irradiation history, and fuel design. Coolant conditions of pressure, temperature, and flow rate that are typical of hot-startup conditions in commercial boiling water reactors (BWRs) will be used. The first test in Series I, Test RIA 1-1, will be comprised of four individual rods, each surrounded by a separate flow shroud. Two rods will be preirradiated and two rods will be unirradiated. The specific objectives of the test are to: (1) characterize the response of unirradiated and preirradiated fuel rods during a RIA event conducted at BWR hot-startup conditions and (2) evaluate test instrumentation response during an RIA. The test sequence will begin with steady state power operation to condition the fuel (pellet cracking and relocation) and determine the fuel rod power calibration. The loop will then be cooled down, the test train removed from the in-pile tube, and one of the unirradiated rods will be removed for fission product analysis and replaced with an identical unirradiated rod. The transient fuel rod energy deposition for Test RIA 1-1 will be chosen from the fuel rod response vs. energy deposition data observed in the first three phases of the RIA Scoping Test. It is anticipated that a fuel pellet surface energy deposition of about 1100 J/g will be required to ensure cladding failure of all four rods. The design of the test fuel rods, test assembly, and instrumentation associated with Test RIA 1-1 are described. The experiment conduct for the test is described. The data recording and reduction requirements are …
Date: August 1, 1978
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY MONTHLY INFORMAL TECHNICAL PROGRESS REPORT SEPTEMBER 1969

Description: This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during September 1969.
Date: October 7, 1969
Creator: Astely, E. R. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department

LOCA Simulation in the National Research Universal Reactor Program Postirradiation Examination Results for the Third Materials Experiment (MT-3) - Second Campaign

Description: A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR} fuel bundles as part of the Loss-of-Coolant Accident (LOCA} Simulation Program by Pacific Northwest Laboratory (PNL). The third materials test (MT-3} was the sixth experiment in a series of thermalhydraulic and materials deformation/rupture experiments conducted in the National Research Universal (NRU) Reactor, Chalk River, Ontario, Canada. The MT-3 experiment was jointly funded by the U.S. Nuclear Regulatory Commission (NRC) and the United Kingdom Atomic Energy Authority (UKAEA) with the main objective of evaluating ballooning and rupture during active two-phase cooling at elevated temperatures. All 12 test rods in the center of the 32-rod bundle failed with an average peak strain of 55.4%. At the request of the UKAEA, a destructive postirradiation examination (PIE) was performed on 7 of the 12 test rods. The results of this examination were presented in a previous report. Subsequently, and at the request of UKAEA, PIE was performed on three additional rods along with further examination of one of the previously examined rods. Information obtained from the PIE included cladding thickness measurements, cladding metallography, and particle size analysis of the fractured fuel pellets. This report describes the additional PIE work performed and presents the results of the examinations.
Date: June 1, 1985
Creator: Haberman, J. H.
Partner: UNT Libraries Government Documents Department

RIA 1-2 Experiment Specification Document

Description: The current status of Reactivity Initiated Accident (RIA) irradiated fuel behavior knowledge, based primarily on the review of the NSRR and SPERT data and the fuel rod modeling studies presented in the Experiment Requirements Document (ERD), is discussed. The hardware specifications are provided, including the test assembly and the test rods. The measurement requirements needed to meet the objectives of the test are presented. Some of the reactor operation requirements associated with the RIA 1-2 test are listed.
Date: July 1, 1977
Creator: Martinson, Z. R. & Eaton, A. M.
Partner: UNT Libraries Government Documents Department

Final Physics Report for the Engineering Test Reactor

Description: This volume continues with discussions of shielding provided for the heat exchanger building, concrete biological shield, top area and bottom area shielding, canal shielding, water shielding requirements during fuel element exchanges, and supplementary shielding requirements.
Date: June 25, 1956
Creator: Cegelski, W. & Machell, W.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY MONTHLY INFORMAL TECHNICAL PROGRESS REPORT OCTOBER 1969

Description: This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during October 1969.
Date: November 7, 1969
Creator: Astely, E. R. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department

Final Physics Report for the Engineering Test Reactor

Description: This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black �control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac …
Date: June 25, 1956
Creator: Wolfe, I. B.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY PERIODIC TECHNICAL REPORT JULY-DECEMBER 1967

Description: Work performed by Battelle-Northwest and its supporting contractors on the AEC-sponsored Fast Flux Test Facility Project is summarized and interpreted for the reporting period. Project activities are reported herein under the following major headings: Plant Design, Construction, and Operation; Components Development; Instrument and Controls Development; Sodium Technology, Core Development; Reactor Materials Development; Fuels Development; Physics; and Reactor Safeguards.
Date: March 1, 1969
Creator: Cabell, C. P.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL SYSTEM DESIGN DESCRIPTION FOR THE PLANT FIRE PROTECTION SYSTEM No. 26

Description: The fire protection system for the FFTF is required for the safety of personnel and the protection of property. Included are functions and design requirements, a detailed description of the system, safety considerations, principles of operation, and maintenance principles.
Date: August 20, 1968
Partner: UNT Libraries Government Documents Department

HIGH TEMPERATURE MODERATOR PROGRAM

Description: The purpose of this memorandum is to outline the high temperature hydride moderator program proposed for the.Metallurgy Division. The objectives of this program are (1) to provide physical and mechanical property data required by the reactor designers, (2) to develop methods for fabricating moderator assemblies, and (3) to devise.and conduct tests to evaluate these· assemblies. The requirements in each of these areas and the work proposed to meet them are outlined.
Date: June 12, 1957
Creator: Hikido, T.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY QUARETLY TECHNICAL REPORT SEPTEMBER, OCTOBER, NOVEMBER 1969

Description: This report was prepared by Battelle-Northwest under Contract No. AT(45-1) -1830 for the Atomic Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during September, October and November 1969.
Date: January 1, 1970
Creator: Astely, E. R. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department

REACTIVITY INITIATED ACCIDENT TEST SERIES TEST RIA 1-4 EXPERIMENT PREDICTIONS

Description: The results of the pretest analyses for Test RIA 1-4 are presented. Test RIA 1-4 consists of a 3x3 array of previously irradiated MAP! fuel rods. The rods have 5.7% enriched UO{sub 2} fuel in zircaloy-4 cladding with an average burnup of 5300 MWd/t. The objective for Test RIA 1-4 is to provide information regarding loss-of-coolable fuel rod geometry following RIA event for a radial-average peak fuel enthalpy equivalent to the present licensing criteria of 1172 J/g (280 cal/g UO{sub 2}). Radial averaged peak fuel enthalpies of 1172 J/g (280 cal/g) 1077 J/g {257 cal/g), and 978 J/g (234 cal/g) for the corner, side, and center fuel rods, respectively, are planned to be achieved during a 2.7 ms reactor period power burst. The results of the FRAP-T5 analyses indicate that all nine rods will fail within 26 ms from the start of the power burst due to pellet-cladding mechanical interaction. All of the rods will undergo partial fuel melting. All rods will operate under extended film boiling (>30 sec) conditions and about 70% of the cladding length is expected to be molten. Approximately 15% of the cladding thickness will be oxided. Fuel swelling due to fission gas release and melting combined with fuel and cladding fragmentation, will probably produce a complete coolant flow blockage within the flow shroud.
Date: February 1980
Creator: Fukuda, S. K. & Martinson, Z. R.
Partner: UNT Libraries Government Documents Department

Blanket Biological Review for General Maintenance Activities Within Active Burial Grounds, 200 E and 200 W Areas, ECR #2000-200-013

Description: No plant and animal species protected under the ESA, candidates for such protection, or species listed by the Washington state government were observed in the vicinity of the proposed sites. Piper's daisy may still occur in some of the burial grounds. This is a Washington State Sensitive plant species, and as such is a Level III resource under the Hanford Site Biological Resources Management Plan. Compensatory mitigation is appropriate for this species when adverse impacts cannot be avoided. The Ecological Compliance Assessment Project (ECAP) staff should consulted prior to the initiation of major work activities within areas where this species has been identified (218-E-12, 218-E-10). The stalked-pod and crouching milkvetch are relatively common throughout 200 West area, therefore even if the few individuals within the active burial grounds are disturbed, it is not likely that the overall local population will be adversely affected. The Watch List is the lowest level of listing for plant species of concern in the State of Washington. No adverse impacts to species or habitats of concern are expected to occur from routine maintenance within the active portions of the 218-W-4C, 218-W-4B, 218-W-3, 218-W-3A, and 218-W-5 burial grounds, as well as the portion of 218-E-12B currently used for storage of retired submarine reactor cores. The remaining portions of the 218-E-12B burial ground, the entire 218-E-10 burial ground, and the 218-W-6 burial ground currently have extensive vegetative cover and it is highly likely that migratory birds, such as meadow larks, horned larks, and curlews will nest in these areas. Therefore, it is recommended that if removal of the existing vegetation is required for burial ground operations, such removal only occur during the August through March time period (i.e. when the birds are not actively nesting). This blanket review does not apply to the portions of 218-W-4C, and …
Date: April 4, 2002
Creator: Sackschewsky, Michael R.
Partner: UNT Libraries Government Documents Department
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