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Evaluation of the effect of break nozzle configuration in loss-of-coolant accident analysis

Description: The Semiscale Mod-1 test program has utilized two different break nozzle configurations in a test facility with identical initial and boundary conditions. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200% double-ended cold leg break loss-of-coolant accident simulation. The first nozzle had a convergent-divergent design; the second nozzle had a convergent design with an elongated constant area throat followed by a rapid expansion. Analysis of data from tests conducted with the two nozzles shows that the critical flow characteristics at the break plane were affected by the break nozzle geometry. Differences in break flow caused differences in the core inlet flow which in turn affected core heater rod thermal response. The results of this investigation show that the break flow behavior and the resulting core thermal response in the Semiscale experimental facility can be directly correlated.
Date: January 1, 1979
Creator: Hanson, R G
Partner: UNT Libraries Government Documents Department

Dry scram evaluation

Description: The analysis performed by Todd Shipyards concerning the ability of the LOFT CRDMs to withstand a dry scram is presented. A ''dry scram'' could result in the CRDM components yielding; however, it would probably not render the CRDMs inoperable. It also concluded that a dry scram is highly unlikely based on a typical LOFT depressurization curve and the temperature of the fluid in the upper pressure housing. At the time of scram, the fluid in the upper pressure housing will not flash to steam owing to the pressure-temperature relationship existing during the scram cycle.
Date: June 6, 1978
Partner: UNT Libraries Government Documents Department

Idaho National Engineering Laboratory: Annual report, 1986

Description: The INEL underwent a year of transition in 1986. Success with new business initiatives, the prospects of even better things to come, and increased national recognition provided the INEL with a glimpse of its promising and exciting future. Among the highlights were: selection of the INEL as the preferred site for the Special Isotope Separation Facility (SIS); the first shipments of core debris from the Three Mile Island Unit 2 reactor to the INEL; dedication of three new facilities - the Fluorinel Dissolution Process, the Remote Analytical Laboratory, and the Stored Waste Experimental Pilot Plant; groundbreaking for the Fuel Processing Restoration Facility; and the first IR-100 award won by the INEL, given for an innovative machine vision system. The INEL has been assigned project management responsibility for the SDI Office-sponsored Multimegawatt Space Reactor and the Air Force-sponsored Multimegawatt Terrestrial Power Plant Project. New Department of Defense initiatives have been realized in projects involving development of prototype defense electronics systems, materials research, and hazardous waste technology. While some of our major reactor safety research programs have been completed, the INEL continues as a leader in advanced reactor technologies development. In April, successful tests were conducted for the development of the Integral Fast Reactor. Other 1986 highlights included the INEL's increased support to the Office of Civilian Radioactive Waste Management for complying with the Nuclear Waste Policy Act of 1982. Major INEL activities included managing a cask procurement program, demonstrating fuel assembly consolidation, and testing spent fuel storage casks. In addition, the INEL supplied the Tennessee Valley Authority with management and personnel experienced in reactor technology, increased basic research programs at the Idaho Research Center, and made numerous outreach efforts to assist the economies of Idaho communities.
Date: January 1, 1986
Partner: UNT Libraries Government Documents Department

Fatigue life of LOFT reactor vessel components

Description: This report provides calculated LOCE lifetimes of the LOFT reactor vessel components. For each component, lifetime is defined as the greatest number of LOCE transient cycles and/or other non-LOCE transient cycles such that cumulative usage factor does not exceed 1.0. The results are presented in tabular form. The results indicate that transients other than LOCE transients govern with or instead of LOCE transients for some components. The other transients are shown in the results.
Date: October 31, 1978
Creator: Kido, C. & Murdock, S.M.
Partner: UNT Libraries Government Documents Department

High rate 4. pi. beta. -. gamma. coincidence counting system

Description: A high count rate 4..pi.. ..beta..-..gamma.. coincidence counting system for the determination of absolute disintegration rates of short half-life radionuclides is described. With this system the dead time per pulse is minimized by not stretching any pulses beyond the width necessary to satisfy overlap coincidence requirements. The equations used to correct for the ..beta.., ..gamma.., and coincidence channel dead times and for accidental coincidences are presented but not rigorously developed. Experimental results are presented for a decaying source of /sup 56/Mn initially at 2 x 10/sup 6/ d/s and a set of /sup 60/Co sources of accurately known source strengths varying from 10/sup 3/ to 2 x 10/sup 6/ d/s. A check of the accidental coincidence equation for the case of two independent sources with varying source strengths is presented.
Date: January 1, 1978
Creator: Johnson, L.O. & Gehrke, R.J.
Partner: UNT Libraries Government Documents Department

Development of process and storage materials suitable for krypton-85 waste management

Description: Screening tests for eight materials (4130 steel, 304 stainless steel, 316 stainless steel, 347 stainless steel, nitronic 50, A286, Monel 400, and Inconel 600) in high purity rubidium showed no liquid metal embrittlement for statically stressed, smooth ''C''-rings at temperatures between 400 and 672/sup 0/K. Potentially injurious localized corrosion in the form of pitting was observed for 304 stainless steel at 672/sup 0/K. All other materials showed good performance in the temperature range proposed for krypton-85 gas cylinder storage. Type 304 stainless steel showed significant general corrosion and intergranular attack in liquid rubidium of lower purity between 793 and 893/sup 0/K. Type 304 stainless steel is not recommended for hardware which may encounter similar service conditions.
Date: January 1, 1978
Creator: Pinchback, T.R. & Knecht, D.A.
Partner: UNT Libraries Government Documents Department

Fluid loads on LOFT DTT shrouds located in reactor vessel downcomer and DTT thermal loads during nuclear LOCE

Description: A thermal analysis was performed on a LOFT drag disc turbine (DTT) for a nuclear LOCE. Thermal gradients through the DTT shroud and turbine body, and temperature differences between body and shroud were calculated. This was done to determine the possible need for additional stress analyses of the DTT shroud to body welds based on thermal loads for a nuclear LOCE.
Date: May 2, 1978
Creator: Kyllingstad, G.
Partner: UNT Libraries Government Documents Department

Fuel behavior during a LOCA: LOFT experiments

Description: The LOFT experiments have provided the following fuel behavior information which appears to be valuable for improving the safety of PWR operation and resolving PWR licensing issues: (1) A generic unassisted core cooling event occurs during large-break LOCAs that dominates the cooling of the core before ECC reflood commences and potentially eliminates the possibility of flow channel blockage from prepressurized fuel rod swelling. (2) The large-break LOCA decompression forces do not disturb the normal control rod gravity drop and may not structually damage the fuel assemblies. (3) Large-break LOCA core cooling may also be enhanced by spacer grid and core counter flow delay of liquid escape from the core boundaries and liquid fallback from the upper plenum into the core region. (4) Lower fuel rod prepressurization may be possible in PWR fuel rods which would reduce flow channel blockage complications during LOCA's. (5) Uniform fuel rod cladding temperature indications during the large break LOCA's do not confirm expectations for the fuel rod cladding temperature variations that would inhibit development of flow channel blockages by ballooning of prepressurized fuel rods.
Date: November 1, 1980
Creator: Russell, M.L.
Partner: UNT Libraries Government Documents Department

Experience with soluble neutron poisons for criticality control at ICPP

Description: Soluble neutron poisons assure criticality control in two of the headend fuel reprocessing systems at the Idaho Chemical Processing Plant. Soluble poisons have been used successfully since 1964 and will be employed in the projected new headend processes. The use of soluble poisons (1) greatly increases the process output (2) allows versatility in the size of fuel assemblies processed and (3) allows the practical reprocessing of some fuels. The safety limit for all fluids entering the U-Zr alloy dissolver is 3.6 g/liter boron. To allow for possible deviations in the measurement systems and drift between analytical sampling periods, the standard practice is to use 3.85 g/liter boron as the lower limit. This dissolver has had 4000 successful hours of operation using soluble poisons. The electrolytic dissolution process depends on soluble gadolinium for criticality safety. This system is used to process high enriched uranium clad in stainless steel. Electrolytic dissolution takes advantage of the anodic corrosion that occurs when a large electrical current is passed through the fuel elements in a corrosive environment. Three control methods are used on each headend system. First, the poison is mixed according to standard operating procedures and the measurements are affirmed by the operator's supervisor. Second, the poisoned solution is stirred, sampled, analyzed, and the analysis reported while still in the mix tank. Finally, a Nuclear Poison Detection System (NPDS) must show an acceptable poison concentration before the solution can be transferred. The major disadvantage of using soluble poisons is the need for very sophisticated control systems and procedures, which require extensive checkout. The need for a poisoned primary heating and cooling system means a secondary system is needed as well. Experience has shown, however, that production enhancement more than makes up for the problems.
Date: January 1, 1978
Creator: Wilson, R.E. & Mortimer, S.R.
Partner: UNT Libraries Government Documents Department

Containment accident analysis using CONTEMPT4/M0D2 compared with experimental data. [PWR]

Description: CONTEMPT4/MOD2 is a new computer program developed to predict the long-term thermal hydraulic behavior of light-water reactor and experimental containment systems during postulated loss-of-coolant accident (LOCA) conditions. Improvements over previous containment codes include multicompartment capability and ice condenser analytical models. A program description and comparisons of calculated results with experimental data are presented.
Date: January 1, 1978
Creator: Metcalfe, L.J.; Hargroves, D.W. & Wells, R.A.
Partner: UNT Libraries Government Documents Department

Considerations of scaling effects in the LOFT reactor system during a large break LOCA simulation

Description: An investigation was performed to assess the effects of scale in a reduced-scale integral test facility designed to simulate the response of a commercial four-loop pressurized water reactor (PWR) during a hypothesized loss-of-coolant accident (LOCA). The facility considered in the investigation was the Loss-of-Fluid Test (LOFT) system, which simulates the principal physical features of a PWR, but has only one-fiftieth of the fluid volume. LOFT experimental data and data from comparable Semiscale Mod-1 and Mod-3 tests were used to assess the influences of component scaling characteristics on LOFT performance during a 200% cold leg break LOCA simulation.
Date: January 1, 1979
Creator: Langerman, M.A. & Harvego, E.A.
Partner: UNT Libraries Government Documents Department

Evaluation of alternate LOFT secondary heat removal systems

Description: The first system proposed for test in the LOFT facility is a pressurized water reactor system, designed to operate continuously for up to 1600 effective full power hours at 50 MWt. Because of the experimental nature of the plant and its relatively short life, utilization of the reactor heat for power generation is not practical. Consequently, it is necessary to provide a secondary heat removal system capable of continuously removing up to 170 x 10/sup 6/ Btu/hr (50 MWt) from the primary coolant system and dissipating this heat to the atmosphere. In the LOFT Conceptual Design, IDO-16833 (Revision I), a heat removal system was proposed in which liquid ''Dowtherm A'' is circulated in a closed secondary loop between the primary heat exchanger and shell-and-tube secondary heat exchangers; in the latter the heat is in turn transferred to a tertiary system in which treated water is circulated between the secondary heat exchangers and an induced draft cooling tower. In response to a KE recommendation that certain alternate heat removal systems might result in cost savings or improved operation, the Idaho Operations Office, USAEC, authorized an engineering and economic study comparing the Dowtherm system described above with a secondary system in which steam is generated from demineralized water in the shell side of the primary heat exchanger, the steam condensed and the condensate recirculated back to the primary heat exchanger in a closed loop. This steam generation concept was to be evaluated both using a water-cooled condenser (the water circulating to a cooling tower as proposed in the Dowtherm system), and using an air-cooled condenser, eliminating the tertiary system. The results of this study are presented.
Date: May 9, 1978
Creator: Stevenson, D.H.
Partner: UNT Libraries Government Documents Department

Electrical, instrumentation, and control codes and standards

Description: During recent years numerous documents in the form of codes and standards have been developed and published to provide design, fabrication and construction rules and criteria applicable to instrumentation, control and power distribution facilities for nuclear power plants. The contents of this LTR were prepared by NUS Corporation under Subcontract K5108 and provide a consolidated index and listing of the documents selected for their application to procurement of materials and design of modifications and new construction at the LOFT facility. These codes and standards should be applied together with the National Electrical Code, the ID Engineering Standards and LOFT Specifications to all LOFT instrument and electrical design activities.
Date: June 7, 1978
Creator: Kranning, A.N.
Partner: UNT Libraries Government Documents Department

Effects of radiation on TFTR coil materials

Description: The coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory was used to irradiate coil insulation specimens provided by the Princeton Plasma Physics Laboratory. Exposure times were chosen as to be representative of expected lifetime doses in the TFTR. Shear bond and flexure tests were performed on irradiated samples; identical unirridated samples were tested as controls. A general loss of strength (10 to 20%) was seen on the majority of specimens.
Date: January 1, 1979
Creator: Imel, G.R.; Kelsey, P.V. & Ottewitte, E.H.
Partner: UNT Libraries Government Documents Department

Dynamic analysis of LOFT reactor flow skirt/core filler assembly for LOCA + SSE

Description: A detailed dynamic analysis of the LOFT reactor core support structures was performed to determine the ability of the flow skirt/core filler and hold-down springs to withstand Loss-of-Coolant Accident (LOCA) plus Safe Shutdown Earthquake (SSE) loadings. A double-ended offset shear occurring in 15 msec (5 msec break time + msec for offset to occur) in the intact loop at the reactor vessel nozzle provided the basis for LOCA loads. The flow skirt/core filler and lower core support structure separate from the core barrel approximately 0.068 in. as a result of the hot leg LOCA. This small displacement and the resulting impact loads produce stresses in the springs, core barrel, flow skirt/core filler, and shear pins within allowables as specified in Section III of the ASME Code for faulted conditions.
Date: June 6, 1978
Creator: Blandford, R.K.
Partner: UNT Libraries Government Documents Department

Fracture toughness of irradiated beryllium. [Fast neutron irradiation]

Description: The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10/sup 21/ n/cm/sup 2/, E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup /sup 1///sub 2//, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup /sup 1///sub 2//, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed.
Date: January 1, 1978
Creator: Beeston, J.M.
Partner: UNT Libraries Government Documents Department

Fracture mechanics evaluation of some LOFT blowdown system and primary coolant coldleg welds

Description: Fracture mechanics evaluations were performed for three welds in the LOFT blowdown system and one weld in the LOFT primary coolant system. Because the applied stress is not known, a sensitivity analysis was run. The assumed initial defect size was one that had a small probability of being missed; applied stresses of 68.9, 137.8, 206.7, and 344.8 MPa were used. It was found that at the lowest stress (68.9 MPa or 10 ksi) the number of cycles from the initial size to rupture was over 6 x 10/sup 6/. The current calculations indicate that with the worst crack configuration--depth-to-length ratio (a/2c) of 0.10--about 1000 cycles with a peak stress of 227.5 MPa (33 ksi) will be needed to propagate the 0.5 x 5.1 cm (0.2 x 2.0 in.) crack to failure.
Date: May 2, 1978
Creator: Nagata, P. K.
Partner: UNT Libraries Government Documents Department

Implementation of a nonequilibrium condensation model in RELAP4/MOD7

Description: RALAp, which is used to simulate the thermal hydraulic behavior of light water reactors subjected to various LOCA transients, is based on the assumption of thermodynamic equilibrium between liquid and vapor within fluid volumes. This assumption, while being appropriate for much of a LOCA transient, is not adequate during the ECC accumulator injection phase as determined by comparisons of code calculations with experimental data. To overcome this limitation, a general model to simulate the nonequilibrium phenomena associated with the mixing of subcooled water with saturated steam has been developed and is operational on preliminary versions of RELAP4/MOD7.
Date: January 1, 1979
Creator: Fischer, S.R.; Chow, H. & Van Arsdall, G.
Partner: UNT Libraries Government Documents Department

INEL Geothermal Environmental Program. Final environmental report

Description: An overview of environmental monitoring programs and research during development of a moderate temperature geothermal resource in the Raft River Valley is presented. One of the major objectives was to develop programs for environmental assessment and protection that could serve as an example for similar types of development. The monitoring studies were designed to establish baseline conditions (predevelopment) of the physical, biological, and human environment. Potential changes were assessed and adverse environmental impacts minimized. No major environmental impacts resulted from development of the Raft River Geothermal Research Facility. The results of the physical, biological, and human environment monitoring programs are summarized.
Date: September 1, 1982
Creator: Thurow, T.L. & Cahn, L.S.
Partner: UNT Libraries Government Documents Department

Loft waste gas processing system analysis. Line 1-1/2''-BSV-10-VD outside penetration S-5C

Description: Lines 1''-BSV-10-VD and 1/sup 1///sub 2/''-BSV-10-VD were analyzed to ASME Section III, Subsection NC (Class 2) criteria. The lines are part of the Waste Gas Processing System. The model considered the portion of the lines outside the containment. The analysis showed that Class 2 requirements will be met with the pipe supports specified in this report.
Date: August 16, 1978
Creator: Pierce, A.T.
Partner: UNT Libraries Government Documents Department

LOFT steady state critical heat flux tests (6. 9 to 13. 8 MPa)

Description: Steady state critical heat flux (CHF) tests have been performed on electrically heated rod bundles simulating the central region of the Loss-of-Fluid Test (LOFT) nuclear reactor core. Previously reported steady state CHF tests have shown that cladding surface thermocouples on LOFT fuel rods reduce the critical heat flux over the pressure range of 13.8 to 16.5 MPa. Reported are additional steady state CHF tests which have been performed to determine the effects of rod external thermocouples on CHF over the ranges of pressure (about 11 MPa) and quality (30 to 40 percent) where CHF is predicted to occur in LOFT during blowdown operation and to determine if sufficient data could be obtained to develop a CHF correlation to predict critical heat fluxes in the LOFT core for blowdown operation.
Date: June 1, 1978
Creator: Gottula, R.C.
Partner: UNT Libraries Government Documents Department

LOFT two-phase flow data integrity analysis

Description: Data integrity methods have been developed and applied to Loss-Of-Fluid Test (LOFT) nuclear reactor safety experiments at the Idaho National Engineering Laboratory. The methods for imroving and qualifying the accuracy of transient measurements on complex thermal-hydraulic experiments are described. These methods involve use of all the information available on the transducers, including data taken during the LOFT experiment itself. Optimum use of these methods determine, in part, the instrumentation package provided on the experiment.
Date: January 1, 1979
Creator: Goodrich, L.D. & Leach, L.P.
Partner: UNT Libraries Government Documents Department