Replaceable LMFBR core components

International Nuclear Information System (INIS)

Evans, E.A.; Cunningham, G.W.

1976-01-01

Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

Validation of SSC using the FFTF natural-circulation tests

International Nuclear Information System (INIS)

Horak, W.C.; Guppy, J.G.; Kennett, R.J.

1982-01-01

As part of the Super System Code (SSC) validation program, the 100% power FFTF natural circulation test has been simulated using SSC. A detailed 19 channel, 2 loop model was used in SSC. Comparisons showed SSC calculations to be in good agreement with the Fast Flux Test Facility (FFTF), test data. Simulation of the test was obtained in real time

Testing experience with fast flux test facility

International Nuclear Information System (INIS)

Noordhoff, B.H.; McGough, C.B.; Nolan, J.E.

1975-01-01

Early FFTF project planning emphasized partial and full-scale testing of major reactor and plant prototype components under expected environmental conditions, excluding radiation fields. Confirmation of component performance during FFTF service was considered essential before actual FFTF startup, to provide increased assurance against FFTF startup delays or operational difficulties and downtime. Several new sodium facilities were constructed, and confirmation tests on the prototype components are now in progress. Test conditions and results to date are reported for the primary pump, intermediate heat exchanger, sodium-to-air dump heat exchanger, large and small sodium valves, purification cold trap, in-vessel handling machine, instrument tree, core restraint, control rod system, low-level flux monitor, closed loop ex-vessel machine, refueling equipment, and selected maintenance equipment. The significance and contribution of these tests to the FFTF and Liquid Metal Fast Breeder Reactor (LMFBR) program are summarized. (U.S.)

Reusable fuel test assembly for the FFTF

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1992-01-01

A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies

Test reactor: basic to U.S. breeder reactor development

International Nuclear Information System (INIS)

Miller, B.J.; Harness, A.J.

1975-01-01

Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

Results of the 1986 FFTF inherent safety tests

International Nuclear Information System (INIS)

Burke, T.M.; Campbell, L.R.; Franz, G.R.; Knecht, W.L.

1987-01-01

A series of tests was recently completed at the 400-MW (thermal) Fast Flux Test Facility (FFTF) to further demonstrate the passive safety characteristics of liquid-metal-cooled fast reactors. Earlier FFTF testing of decay heat removal by sodium natural circulation was reported in 1981. The main purpose of the 1986 test series was to demonstrate passive reactor shutdown during a loss-of-flow event when several inherent shutdown devices called gas expansion modules (GEMs) were installed in the reactor. However, these tests also provide further data on the natural circulation performance of the primary system, in particular the reactor core, and thus add to the data base available for checking the validity of available analytical tools

Secure Retrieval of FFTF Testing, Design, and Operating Information

International Nuclear Information System (INIS)

Butner, R. Scott; Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah

2009-01-01

One of the goals of the Advanced Fuel Cycle Initiative (AFCI) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). In addition, preserving LMR information and knowledge is part of a larger international collaborative activity conducted under the auspices of the International Atomic Energy Agency (IAEA). A similar program is being conducted for EBR-II at the Idaho Nuclear Laboratory (INL) and international programs are also in progress. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. As the primary function of the FFTF was testing, the focus is also on preserving information obtained from irradiation testing of fuels and materials. This information will be invaluable when, at a later date, international decisions are made to pursue new LMRs. In the interim, this information may be of potential use for international exchanges with other LMR programs around the world. At least as important in the United States, which is emphasizing large-scale computer simulation and modeling, this information provides the basis for creating benchmarks for validating and testing these large scale computer programs. Although the preservation activity with respect to FFTF information as discussed below is still underway, the team of authors above is currently retrieving and providing experimental and design information to the LMR modeling and simulation efforts for use in validating their computer models. On the Hanford Site, the FFTF reactor plant is one of the facilities intended for decontamination and decommissioning consistent with the cleanup mission on this site. The reactor facility has been deactivated and is being maintained in a cold and dark minimal surveillance and maintenance mode until final decommissioning is pursued. In order to ensure protection of information at risk, the program to date has focused on sequestering and secure retrieval

Fast Flux Test Facility (FFTF) Briefing Book 1 Summary

Energy Technology Data Exchange (ETDEWEB)

WJ Apley

1997-12-01

This report documents the results of evaluations preformed during 1997 to determine what, if an, future role the Fast Flux Test Facility (FFTF) might have in support of the Department of Energy’s tritium productions strategy. An evaluation was also conducted to assess the potential for the FFTF to produce medical isotopes. No safety, environmental, or technical issues associated with producing 1.5 kilograms of tritium per year in the FFTF have been identified that would change the previous evaluations by the Department of Energy, the JASON panel, or Putnam, Hayes & Bartlett. The FFTF can be refitted and restated by July 2002 for a total expenditure of $371 million, with an additional $64 million of startup expense necessary to incorporate the production of medical isotopes. Therapeutic and diagnostic applications of reactor-generated medical isotopes will increase dramatically over the next decade. Essential medical isotopes can be produced in the FFTF simultaneously with tritium production, and while a stand-alone medical isotope mission for the facility cannot be economically justified given current marker conditions, conservative estimates based on a report by Frost &Sullivan indicate that 60% of the annual operational costs (reactor and fuel supply) could be offset by revenues from medical isotope production within 10 yeas of restart. The recommendation of the report is for the Department of Energy to continue to maintain the FFTF in standby and proceed with preparation of appropriate Nations Environmental Policy Act documentation in full consultation with the public to consider the FFTF as an interim tritium production option (1.5 kilograms/year) with a secondary mission of producing medical isotopes.

Analysis of in-core coolant temperatures of FFTF instrumented fuels tests at full power

International Nuclear Information System (INIS)

Hoth, C.W.

1981-01-01

Two full size highly instrumented fuel assemblies were inserted into the core of the Fast Flux Test Facility in December of 1979. The major objectives of these instrumented tests are to provide verification of the FFTF core conditions and to characterize temperature patterns within FFTF driver fuel assemblies. A review is presented of the results obtained during the power ascents and during irradiation at a constant reactor power of 400 MWt. The results obtained from these instrumented tests verify the conservative nature of the design methods used to establish core conditions in FFTF. The success of these tests also demonstrates the ability to design, fabricate, install and irradiate complex, instrumented fuel tests in FFTF using commercially procured components

A new approach to the design of LMFBR liners

International Nuclear Information System (INIS)

Polentz, L.M.

1980-01-01

An advance in the state-of-the-art of LMFBR liners which permits notable savings in construction costs without any sacrifice of safety is described. The application of the new design concept to the rework of the upper reactor vault liner of the FFTF is discussed. Factors which affect the application of the new design approach to other LMFBRs are delineated and discussed. (author)

Fast Flux Test Facility (FFTF) standby plan

Energy Technology Data Exchange (ETDEWEB)

Hulvey, R.K.

1997-03-06

The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy`s dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode.

Fast Flux Test Facility (FFTF) standby plan

International Nuclear Information System (INIS)

Hulvey, R.K.

1997-01-01

The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy's dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode

Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

International Nuclear Information System (INIS)

Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

1979-01-01

A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

Axial migratin of cesium in LMFBR fuel pins

International Nuclear Information System (INIS)

Karnesky, R.A.; Bridges, A.E.; Jost, J.W.

1981-11-01

A correlated model for quantitatively predicting the behavior of cesium in LMFBR fuel pins has been developed. This correlation was shown to be in good agreement with experimental data. It has been used to predict the behavior of cesium in the FFTF driver fuel and as the result of this analysis it has been shown that the accumulation of cesium in the insulator pellets at the ends of the fuel column will not be life limiting

Status of U.S. LMFBR programme

International Nuclear Information System (INIS)

Yevich, J.

1978-01-01

The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

Validation of turbulence models for LMFBR outlet plenum flows

International Nuclear Information System (INIS)

Chen, Y.B.; Golay, M.W.

1977-01-01

Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

FFTF [Fast Flux Test Facility] fuel handling experience (1979--1986)

International Nuclear Information System (INIS)

Romrell, D.M.; Art, D.M.; Redekopp, R.D.; Waldo, J.B.

1987-05-01

The Fast Flux Test Facility (FFTF)is a 400 MW (th) sodium-cooled fast flux test reactor located on the Hanford Site in southeastern Washington State. The FFTF is operated by the Westinghouse Hanford Company for the United States Department of Energy. The FFTF is a three loop plant designed primarily for the purpose of testing full-scale core components in an environment prototypic of future liquid metal reactors. The plant design emphasizes features to enhance this test capability, especially in the area of the core, reactor vessel, and refueling system. Eight special test positions are provided in the vessel head to permit contact instrumented experiments to be installed and irradiated. These test positions effectively divide the core into three sectors. Each sector requires its own In-Vessel Handling Machine (IVHM) to access all the core positions. Since the core and the in-vessel refueling components are submerged under sodium, all handling operations must be performed blind. This puts severe requirements on the positioning ability are reliability of the refueling components. This report addresses the operating experience with the fuel handling system from initial core loading in November, 1979 through 1986. This includes 9 refueling cycles. 2 refs., 8 figs

Transient testing of the FFTF for decay-heat removal by natural convection

International Nuclear Information System (INIS)

Beaver, T.R.; Johnson, H.G.; Stover, R.L.

1982-06-01

This paper reports on the series of transient tests performed in the FFTF as a major part of the pre-operations testing program. The structure of the transient test program was designed to verify the capability of the FFTF to safely remove decay heat by natural convection. The series culminated in a scram from full power to complete natural convection in the plant, simulating a loss of all electrical power. Test results and acceptance criteria related to the verification of safe decay heat removal are presented

Application of mass spectrometry to fuels and materials testing at FFTF

International Nuclear Information System (INIS)

Plucinski, C.E.; Goheen, M.W.; McCown, J.J.

1983-01-01

The Fast Flux Test Facility (FFTF) is a sodium cooled reactor operated for the Department of Energy by Westinghouse Hanford Company. In FFTF the 78 fuel pin assemblies have been tagged with a unique combination of Kr and Xe isotopes. In addition to fuel pin assemblies, other test assemblies can also be similarly tagged as in the Materials Open Test Assembly (MOTA). During power operations leaks are monitored by single channel gamma analyzers in the reactor cover gas system. When radioactive Kr and Xe isotopes are detected a gas tag sample trap (GTST) is used to take a sample of the reactor cover gas, the Kr and Xe are concentration, and isotopic analysis is obtained by mass spectrometry

FFTF [Fast Flux Test Facility] management

International Nuclear Information System (INIS)

Bennett, C.L.

1986-11-01

Fuel Management at the Fast Flux Test Facility (FFTF) involves more than just the usual ex-core and in-core management of standard fuel and non-fuel components between storage locations and within the core since it is primarily an irradiation test facility. This mission involves testing an ever increasing variety of fueled and non-fueled experiments, each having unique requirements on the reactor core as well as having its own individual impact on the reload design. This paper describes the fuel management process used by the Westinghouse Hanford Company Core Engineering group that has led to the successful reload design of nine operating cycles and the irradiation of over 120 tests

Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

John D. Bess

2010-03-01

The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

Experimental plans for LMFBR cavity liner sodium spill test LT-1

International Nuclear Information System (INIS)

Hilliard, R.K.; Newell, G.A.

1976-01-01

Reinforced concrete is an important material of construction in LMFBR cavities and cells. Steel liners are often installed on the concrete surfaces to provide a gastight seal for minimizing air inleakage to inerted cell atmospheres and to protect the concrete from direct contact with sodium in the event of a sodium spill. In making safety assessment analyses, it is of interest to determine the adequacy of the liners to maintain their leaktightness during postulated accidents involving large sodium spills. However, data for basing analytical assessments of cell liners are very meager and an experimental program is underway at HEDL to provide some of the needed information. The HEDL cell liner evaluation program consists of both bench-scale feature tests and large-scale sodium spill demonstration tests. The plans for the first large-scale sodium spill test (LT-1) are the subject of this paper

FFTF integrated leak rate computer system

International Nuclear Information System (INIS)

Hubbard, J.A.

1987-01-01

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. The FFTF is the only reactor of this type designed and operated to meet the licensing requirements of the Nuclear Regulatory Commission. Unique characteristics of the FFTF that present special challenges related to leak rate testing include thin wall containment vessel construction, cover gas systems that penetrate containment, and a low-pressure design basis accident. The successful completion of the third FFTF integrated leak rate test 5 days ahead of schedule and 10% under budget was a major achievement for the Westinghouse Hanford Company. The success of this operational safety test was due in large part to a special network (LAN) of three IBM PC/XT computers, which monitored the sensor data, calculated the containment vessel leak rate, and displayed test results. The equipment configuration allowed continuous monitoring of the progress of the test independent of the data acquisition and analysis functions, and it also provided overall improved system reliability by permitting immediate switching to backup computers in the event of equipment failure

Manufacturing and testing experience for FFTF major safety related components

International Nuclear Information System (INIS)

Peckinpaugh, C.L.

1976-01-01

Experience with FFTF Heat Transport System components during design, manufacturing, and prototype testing is dscussed. Specifically the special design features and the results of the testing performed to assure that the designs provide for safe operation are outlined. Particular emphasis is placed on the full size prototype testing programs and the valuable experience gained

Structural material irradiations in FFTF

International Nuclear Information System (INIS)

1985-01-01

Information is presented concerning the Materials Open Test Assembly (MOTA); instrumentation and control system; MOTA neutronic data; pressurized tube specimens; stress-rupture measurements for reactor materials; miniature specimen design; the Interim Examination and Maintenance (IEM) cell at the FFTF; support services; and general information concerning the FFTF

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Fast Flux Test Facility (FFTF) feedback reactivity components

International Nuclear Information System (INIS)

Nguyen, D.H.

1988-04-01

The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

Initial acceptance test experience with FFTF plant equipment

International Nuclear Information System (INIS)

Brown, R.K.; Coleman, K.A.; Mahaffey, M.K.; McCargar, C.G.; Young, M.W.

1978-09-01

The purpose of this paper is to examine the initial acceptance test experience of certain pieces of auxiliary equipment of the Fast Flux Test Facility (FFTF). The scope focuses on the DHX blowers and drive train, inert gas blowers, H and V containment isolation valves, and the Surveillance and In-service Inspection (SISI) transporter and trolley. For each type of equipment, the discussion includes a summary of the design and system function, installation history, preoperational acceptance testing procedures and results, and unusual events and resolutions

Design review report FFTF interim storage cask

International Nuclear Information System (INIS)

Scott, P.L.

1995-01-01

Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location

FFTF Plant transition mission analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-01-01

FFTF (Fast Flux Test Facility) is a 400-MW(t) sodium-cooled, fast flux test reactor at Hanford, designed to test fuels and materials for advanced nuclear power plants; it has no capability for generating electric power. Since a long-term mission could not be found for FFTF, it was placed in standby, and a recommendation was made that it be shut down. Purpose of the FFTF Transition Project is to prepare it for Decontamination and Decommissioning; this will be accomplished by establishing a passively safe and environmentally secure configuration, that can be preserved for several decades. This report presents the results of the mission analysis, which is required by Hanford systems engineering procedures

Sodium fill of FFTF

International Nuclear Information System (INIS)

Waldo, J.B.; Greenwell, R.K.; Keasling, T.A.; Collins, J.R.; Klos, D.B.

1980-02-01

With construction of the Fast Flux Test Facility (FFTF) completed, the first major objective in the startup program was to fill the sodium systems. A sodium fill sequence was developed to match construction completion, and as systems became available, they were inerted, preheated, and filled with sodium. The secondary sodium systems were filled first while dry refueling system testing was in progress in the reactor vessel. The reactor vessel and the primary loops were filled last. This paper describes the methods used and some of the key results achieved for this major FFTF objective

Eddy Current Flow Measurements in the FFTF

Energy Technology Data Exchange (ETDEWEB)

Nielsen, Deborah L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Polzin, David L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Makenas, Bruce J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuable information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.

Optimized FFTF Acceptance Test Program covering Phases III, IV, and V

International Nuclear Information System (INIS)

Wykoff, W.R.; Jones, D.H.

1977-03-01

A detailed review of Phases III, IV, and V of the FFTF Acceptance Test Program has been completed. The purpose of this review was to formulate that test sequence which not only meets requirements for safe, reliable and useful operation of the plant, but also results in the earliest prudent demonstration of full-power performance. A test sequence based on the underlying assumption that sodium flows into the secondary sodium storage tank (T-44) no later than August 31, 1978, is described in detail. A time-scale which allows extra time to put systems and equipment into operation the first time, debugging, and learning how to operate most effectively has been superimposed on the test sequence. Time is not included for major equipment malfunctions. This test plan provides the basis for coordinating the many and varied activities and interfaces necessary for successful and timely execution of the FFTF Acceptance Test Program. In this report, the need dates have been identified for presently scheduled test articles and standard core components

Operational safety at the FFTF

International Nuclear Information System (INIS)

Baird, Q.L.; Hagan, J.W.; Seeman, S.E.; Baker, S.M.

1981-02-01

An extensive operational nuclear safety program has been an integral part of the design, startup, and initial operating phases of the Fast Flux Test Facility (FFTF). During the design and construction of the facility, a program of independent safety overviews and analyses assured the provision of responsible safety margins within the plant, protective systems, and engineered safety features for protection of the public, operating staff, and the facility. The program is continuing through surveillance of operations to verify continued adherence to the established operating envelope and for timely identification of any trends potentially adverse to those margins. Experience from operation of FFTF is being utilized in the development of enhanced operational nuclear safety aids for application in follow-on breeder reactor power systems. The commendable plant and personnel safety experiences of FFTF through its startup and ascension to full power demonstrate the overall effectiveness of the FFTF operational nuclear safety program

Fission and corrosion products behavior in primary circuits of LMFBR's

International Nuclear Information System (INIS)

Feuerstein, H.; Thorley, A.W.

1987-08-01

Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

Reactivity anomalies in the FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Knutson, B.J.; Harris, R.A.

1987-04-01

Experience using an automated core reactivity monitoring technique at the Fast Flux Test Facility (FFTF) through eight operating cycles is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. Reactivity worth predictions of core state changes (e.g., temperature and irradiation changes) and compensating control rod movements are required for the rod position comparison. A substantial data base now exists to evaluate changes in temperature reactivity feedback effects operational in the FFTF, rod worth changes due to core loading, temperature and irradiation effects and burnup effects associated with transmutation of fuel materials. This report summarizes preliminary work of correlating zero power and at-power rod worth measurement data, calculated burnup rates and rod worths using the latest ENDF/B-V cross section set for each cycle to evaluate the prediction models and attempt to resolve observed reactivity anomalies. 2 figs., 2 tabs

Remote replacement of materials open-test assembly specimens at the FFTF/IEM cell

International Nuclear Information System (INIS)

Gibbons, P.W.; Ramsey, E.B.

1990-01-01

The Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell is used for the remote disassembly of irradiated fuel and materials experiments. The materials open-test assembly (MOTA) is brought to the IEM cell for materials test specimen removal. The specimens are shipped to the materials laboratory for sorting and installation in new specimen holders and then returned within 10 days to the IEM cell where they are installed in a new MOTA vehicle for further irradiation. Reconstituting a MOTA is a challenging remote operation involving dozens of steps and two separate facilities. Handling and disassembling sodium-wetted components pose interesting handling, cleaning, and disposal challenges. The success of this system is evidenced by its timely completion in the critical path of FFTF outage schedules

Leak testing at Westinghouse Hanford Company for the Fast Flux Test Facility (FFTF)

International Nuclear Information System (INIS)

Jackson, C.N.

1981-01-01

Described leak testing applications require an arsenal of test equipment, a diverse range of testing techniques and a cadre of technical talent. A wide range helium mass spectrometer leak detector, a volume change tester and a halogen detector are employed to cover the 1 x 10 -8 to 1 atm cc/sec leak rate range encountered. Leak testing techniques, equipment problems, costs, and recommendations are discussed for examination of reactor pressure boundary and other ancillary components of the FFTF

Integrated leak rate test of the FFTF [Fast Flux Test Facility] containment vessel

International Nuclear Information System (INIS)

Grygiel, M.L.; Davis, R.H.; Polzin, D.L.; Yule, W.D.

1987-04-01

The third integrated leak rate test (ILRT) performed at the Fast Flux Test Facility (FFTF) demonstrated that effective leak rate measurements could be obtained at a pressure of 2 psig. In addition, innovative data reduction methods demonstrated the ability to accurately account for diurnal variations in containment pressure and temperature. Further development of methods used in this test indicate significant savings in the time and effort required to perform an ILRT on Liquid Metal Reactor Systems with consequent reduction in test costs

Continuous gamma-ray spectrometry in the fast flux test facility (FFTF)

International Nuclear Information System (INIS)

Gold, R.; Kaiser, B.J.; Moore, F.S. Jr.; Bunch, W.L.; McElroy, W.N.; Sheen, E.M.

1980-03-01

In-core Compton-recoil γ-ray spectrometry was carried out in FFTF at very low power in the In-Reactor Thimble (IRT). The lower-energy electron spectrum at the FFTF-IRT midplane and the unfolded γ-ray spectrum are shown. 2 figures

Coolant mixing in the LMFBR outlet plenum

International Nuclear Information System (INIS)

Chen, Y.B.; Golay, M.W.

1977-06-01

Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

Interface agreement for the management of FFTF Spent Nuclear Fuel

International Nuclear Information System (INIS)

McCormack, R.L.

1995-01-01

The Hanford Site Spent Nuclear Fuel (SNF) Project was formed to manage the SNF at Hanford. The mission of the Fast Flux Test Facility (FFTF) Transition Project is to place the facility in a radiologically and industrially safe shutdown condition for turnover to the Environmental Restoration Contractor (ERC) for subsequent D ampersand D. To satisfy both project missions, FFTF SNF must be removed from the FFTF and subsequently dispositioned. This documented provides the interface agreement between FFTF Transition Project and SNF Project for management of the FFTF SNF

Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

International Nuclear Information System (INIS)

Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

1986-01-01

Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

Fast Flux Test Facility

International Nuclear Information System (INIS)

Munn, W.I.

1981-01-01

The Fast Flux Test Facility (FFTF), located on the Hanford site a few miles north of Richland, Washington, is a major link in the chain of development required to sustain and advance Liquid Metal Fast Breeder Reactor (LMFBR) technology in the United States. This 400 MWt sodium cooled reactor is a three loop design, is operated by Westinghouse Hanford Company for the US Department of Energy, and is the largest research reactor of its kind in the world. The purpose of the facility is three-fold: (1) to provide a test bed for components, materials, and breeder reactor fuels which can significantly extend resource reserves; (2) to produce a complete body of base data for the use of liquid sodium in heat transfer systens; and (3) to demonstrate inherent safety characteristics of LMFBR designs

LMFBR intermediate-heat-exchanger experience

International Nuclear Information System (INIS)

Cho, S.M.; Beaver, T.R.

1983-01-01

This paper presents developmental and operating experience of large Intermediate Heat Exchangers (IHX's) in US from the Fast Flux Test Facility (FFTF) to the Clinch River Breeder Reactor Plant (CRBRP) to the Large Development Plant (LDP). Design commonalities and deviations among these IHX's are synopsized. Various developmental tests that were conducted in the areas of hydraulic, structural and mechanical design are also presented. The FFTF is currently operating. Performance data of the FFTF IHXs are reviewed, and comparisons between actual and predicted performances are made. The results are used to assess the adequacy of IHX designs

Seismic response and damping tests of small bore LMFBR piping and supports

International Nuclear Information System (INIS)

Barta, D.A.; Anderson, M.J.; Severud, L.K.; Lindquist, M.R.

1984-01-01

Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

Inference of physical phenomena from FFTF [Fast Flux Test Facility] noise analysis

International Nuclear Information System (INIS)

Thie, J.A.; Damiano, B.; Campbell, L.R.

1989-01-01

The source of features observed in noise spectra collected by an automated data collection system operated by the Oak Ridge National Laboratory at the Fast Flux Test Facility (FFTF) can be identified using a methodology based on careful data observation and intuition. When a large collection of data is available, as in this case, automatic pattern recognition and parameter storage and retrieval using a data base can be used to extract useful information. However, results can be limited to empirical signature comparison monitoring unless an effort is made to determine the noise sources. This paper describes the identification of several FFTF noise data phenomena and suggests how this understanding may lead to new or enhanced monitoring. 13 refs., 4 figs

Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

International Nuclear Information System (INIS)

Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

1982-01-01

LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

Preservation of FFTF Data Related to Passive Safety Testing

International Nuclear Information System (INIS)

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-01-01

One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

FFTF preoperational survey. Program report

International Nuclear Information System (INIS)

Twitty, B.L.; Bicehouse, H.J.

1980-12-01

The FFTF will become operational with criticality early in 1980. This facility is composed of the test reactor, fuel examination cells, expended fuel storage systems and fuel handling systems. The reactor and storage systems are sodium-cooled with the heat load dumped to the ambient air through heat exchangers. In order to assure that the operation of the FFTF has minimal impact on the environment, a monitoring program has been established. Prior to operation of a new facility, a preoperational environmental survey is required. It is the purpose of this report to briefly describe the environmental survey program and to provide the background data obtained during the preoperational phase of the survey program. Nine stations in the program of particular importance to FFTF are discussed in detail with results of monitoring given. No unexplained trends were noted

FFTF operator training

International Nuclear Information System (INIS)

1979-01-01

The qualifications required and the training received by FFTF operators are described. The training includes sodium fill training and cold plant qualification. Requirements for supervisors are also outlined. Arrangement of personnel at FFTF is described. Requalification training and recertification are considered

85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

International Nuclear Information System (INIS)

Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

1983-01-01

The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

Overview of the fast reactors fuels program. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

FFTF Authorization Agreement

International Nuclear Information System (INIS)

DAUTEL, W.A.

2000-01-01

The purpose of the Authorization Agreement is to serve as a mechanism whereby the U.S. Department of Energy, Richland Operations Office (RL) and Fluor Hanford (FH) jointly clarify and agree to key conditions for conducting work safely and efficiently in the Fast Flux Test Facility (FFTF). Work must be accomplished in a manner that achieves high levels of quality while protecting the environment and the safety and health of workers and the public, and complying with applicable contractual and regulatory requirements. It is the intent of this Agreement to address those items of significant importance in establishing and supporting the FFTF Authorization Envelope, but this Agreement in no way alters the terms and conditions of the Project Hanford Management Contract (PHMC), Contract Number DE-AC06-96RL13200

Seismic testing and analysis of a prototypic nonlinear piping system

International Nuclear Information System (INIS)

Barta, D.A.; Anderson, M.J.; Severud, L.K.

1982-11-01

A series of seismic tests and analyses of a nonlinear Fast Flux Test Facility (FFTF) prototypic piping system are described, and measured responses are compared with analytical predictions. The test loop was representative of a typical LMFBR insulated small bore piping system and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps. Various piping support configurations were tested and analyzed to evaluate the effects of free play and other nonlinear stiffness characteristics on the piping system response

FFTF radioactive solid waste handling and transport

International Nuclear Information System (INIS)

Thomson, J.D.

1982-01-01

The equipment necessary for the disposal of radioactive solid waste from the Fast Flux Test Facility (FFTF) is scheduled to be available for operation in late 1982. The plan for disposal of radioactive waste from FFTF will utilize special waste containers, a reusable Solid Waste Cask (SWC) and a Disposable Solid Waste Cask (DSWC). The SWC will be used to transport the waste from the Reactor Containment Building to a concrete and steel DSWC. The DSWC will then be transported to a burial site on the Hanford Reservation near Richland, Washington. Radioactive solid waste generated during the operation of the FFTF consists of activated test assembly hardware, reflectors, in-core shim assemblies and control rods. This radioactive waste must be cleaned (sodium removed) prior to disposal. This paper provides a description of the solid waste disposal process, and the casks and equipment used for handling and transport

FFTF fuel pin design bases and performance

International Nuclear Information System (INIS)

Cox, C.M.; Hanson, J.E.; Roake, W.E.; Slember, R.J.; Weber, C.E.; Millunzi, A.C.

1975-04-01

The FFTF fuel pin was conservatively designed to meet thermal and structural performance requirements in the categories normal operation, upset events, emergency events, and hypothetical, faulted events. The fuel pin operating limits consistent with these requirements were developed from a strong fuel pin irradiation testing program scoped to define the performance capability under relevant steady state and transient conditions. Comparison of the results of the irradiation testing program with design requirements indicates that the FFTF fuel pin can exceed its goal burnup of 80,000 MWd/MTM. (U.S.)

Reactivity shimming in FFTF with safety rods

International Nuclear Information System (INIS)

Bennett, C.L.

1986-01-01

The Fast Flux Test Facility (FFTF) reactor demonstrated operation at a 65% annual capacity factor in 1984 and a 70% annual capacity factor in 1985. This was achieved in part by extending the cycle lengths beyond the originally designed 100-day cycles. The advent of extended cycle lengths at the FFTF reactor led to the development of a contingency plan to shim some of the installed excess reactivity at the beginning of the cycle with the primary control rod bank (safety rods) in order to maintain the minimum required shutdown margin in the secondary shutdown system. This paper describes both the implementation of this plan in terms of the licensing aspects and the actual use of primary shim during cycles 5 and 6 operation at FFTF

Inadvertent raising of levels in the FFTF primary sodium pumps. Final unusual occurrence report, HEDL 79-34 (FFTF-58)

International Nuclear Information System (INIS)

Kuechle, J.D.

1981-01-01

The final unusual occurrence report describes the inadvertent raising of the sodium level in the FFTF primary sodium pumps during system testing. This event is now judged to have caused permanent deformation of the primary pump shaft on loop 1 during a period when pump rotation was stopped and sodium level in the pump tank was inadvertently increased. The shaft was subsequently removed, straightened, and returned to service in the spare FFTF pump

Strain accumulation in a prototypic lmfbr nozzle: Experimental and analytical correlation

International Nuclear Information System (INIS)

Woodward, W.S.; Dhalia, A.K.; Berton, P.A.

1986-01-01

At an early stage in the design of the primary inlet nozzle for the Intermediate Heat Exchanger (IHX) of the Fast Flux Test Facility (FFTF), it was predicted that the inelastic strain accumulation during elevated temperature operation (1050 0 F/566 0 C) would exceed the ASME Code design allowables. Therefore, a proof test of a prototypic FFTF IHX nozzle was performed in the Westinghouse Creep Ratcheting Test Facility (CRTF) to measure the ratchet strain increments during the most severe postulated FFTF plant thermal transients. In addition, analytical procedures similar to those used in the plant design, were used to predict strain accumulation in the CRTF nozzle. This paper describes how the proof test was successfully completed, and it shows that both the test measurements and analytical predictions confirm that the FFTF IHX nozzle, subjected to postulated thermal and mechanical loadings, complies with the ASME Code strain limits. Also, these results provide a measure of validation for the analytical procedures used in the design of FFTF as well as demonstrate the structural adequacy of the FFTF IHX primary inlet nozzle

FFTF startup: status and results

International Nuclear Information System (INIS)

Noordhoff, B.H.; Moore, C.E.

1980-03-01

Startup testing on the Fast Flux Test Facility (FFTF) during the past three years has progressed beyond initial criticality toward the principal goal of power demonstration in 1980. An overview is presented of technical results to date and project plans to achieve power demonstration and complete the startup test program

FFTF reactor assembly system technology

International Nuclear Information System (INIS)

Mangelsdorf, T.A.

1975-01-01

An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

Review of FFTF and CRBRP control rod systems designs

International Nuclear Information System (INIS)

Pitterle, T.A.; Lagally, H.O.

1977-01-01

The evolution of the primary control rod system design for FFTF and CRBR, beginning with the initial choice of the basic concepts, is described. The significant component and systems tests are reviewed together with the test results which referenced the development of the CRBR primary control rod system design. Modifications to the concepts and detail designs of the FFTF control rod system were required principally to satisfy the requirements of CRBR, and at the same time incorporating design refinements shown desirable by the tests

Overview of planning process at FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Gadeken, A.D.

1986-03-01

The planning process at the Fast Flux Test Facility (FFTF) is controlled through a hierarchy of documents ranging from a ten-year strategic plan to a weekly schedule. Within the hierarchy are a Near-Term (three-year) Operating Plan, a Cycle (six-month) Plan, and an Outage/Operating Phase Schedule. Coordination of the planning process is accomplished by a dedicated preparation team that also provides an overview of the formal planning timetable which identifies key action items required to be completed before an outage/operating phase can begin

Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

International Nuclear Information System (INIS)

Stratton, R.W.; Ledergerber, G.; Ingold, F.; Latimer, T.W.; Chidester, K.M.

1993-01-01

The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all t the requirements of the specifications concerning soichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour. (orig.)

Technical specification upgrading at the Fast Flux Test Facility (FFTF)

International Nuclear Information System (INIS)

Baird, Q.L.; Franz, G.R.; Absher, K.R.

1985-01-01

The FFTF Technical Specifications were generated in 1977 and 1978 following submittal of the FSAR in 1976. A phased implementation program served to prepare the specifications for each stage of the plant startup with the complete specifications approved and implemented late in 1980 for the first ascent to full power. In January, 1983 WHC undertook an upgrading effort to implement changes to the FFTF technical specifications. This program has been pursued with appropriate attention to the CFR and industry standards and practice. Examples of these changes, discussion of the methods and planned activities for the future will be presented. Technical data will be provided to support the impact of specific limits. The benefits of changes and the criteria for change will be elaborated

Radiological considerations of the reactor cover gas processing system at the FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Prevo, P.R.

1986-09-01

Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986)

FFTF [Fast Flux Test Facility] performance measurements for safety, productivity and control

International Nuclear Information System (INIS)

Newland, D.J.; Praetorius, P.R.; Tomaszewski, T.A.

1987-05-01

A useful set of performance measurements for Safety, Productivity and Control has evolved at the Fast Flux Test Facility (FFTF). In response to declining budgets and the resulting need to safely manage a manpower rampdown, an ''Early Warning System'' was developed in 1984. Its purpose was to monitor the effects of the staffing rampdown such that appropriate remedial action could be taken to correct adverse trends before a significant problem occurred. 1 tab

Preliminary benefit-cost analysis of the Fast Flux Test Facility (FFTF) power addition

International Nuclear Information System (INIS)

Callaway, J.M.; Lezberg, A.J.; Scott, M.J.; Tawil, J.J.

1984-07-01

The primary objective of this report is to conduct a preliminary benefit-cost study for the proposed power addition to FFTF to determine whether the project is cost-effective. If the project is authorized, construction will begin in 1986 and end in 1991. Full power operation is scheduled to begin in 1991 and a project life of 20 years is assumed. The undiscounted cost during the construction period of the FFTF power addition is estimated to be approximately $117 million over the construction period (1984 dollars). An additional $3 million is estimated as the opportunity cost - or value of these resources in their most favorable alternative use - of surplus FFTF equipment and unused CRBR equipment, including materials for steam generator fabrication. The annual operating and maintenance cost of the project is estimated to be about $2.1 million in 1984 dollars. 20 references

Water tests for determining post voiding behavior in the LMFBR

International Nuclear Information System (INIS)

Hinkle, W.D.

1976-06-01

The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described were conceived with the objective of obtaining basic information and data relating to this possibility

FFTF/IEM cell fuel pin weighing system

International Nuclear Information System (INIS)

Gibbons, P.W.

1987-01-01

The Interim Examination and Maintenance (IEM) cell in the Fast Flux Test Facility (FFTF) is used for remote disassembly of irradiated fuel and materials experiments. For those fuel experiments where the FFTF tag-gas detection system has indicated a fuel pin cladding breach, a weighing system is used in identifying that fuel pin with a reduced weight due to the escape of gaseous and volatile fission products. A fuel pin weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF), was the basis for the IEM cell system. Design modifications to the original equipment were centered around adapting the machine to the differences between the two facilities and correcting deficiencies discovered during functional testing in the IEM cell mock-up

FFTF operating experience with sodium natural circulation: slides included

Energy Technology Data Exchange (ETDEWEB)

Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

1981-01-01

The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures.

FFTF operating experience with sodium natural circulation: slides included

International Nuclear Information System (INIS)

Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

1981-01-01

The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures

Data systems in FFTF and EBR-II

International Nuclear Information System (INIS)

Warrick, R.P.; Ritter, W.M.

1980-02-01

This paper describes the Data System used to monitor operation and collect experimental data in FFTF. This data system has evolved since initial inception from a relatively simple, single computer system monitoring a relatively few (approx. 1000) instrument channels important for operation to one which has increased capability to support the long-range testing needs in FFTF. The system, while still relatively simple, now contains multiple computers which normally perform independent functions. The computers, however, provide backup processing for certain simple tasks. Operator interfacing is provided through CRT's. The output capabilities of the system are described. A description of the Data System in EBR-II is also included

High burnup performance of an advanced oxide fuel assembly in FFTF [Fast Flux Test Facility] with ferritic/martensitic materials

International Nuclear Information System (INIS)

Bridges, A.E.; Saito, G.H.; Lovell, A.J.; Makenas, B.J.

1986-05-01

An advanced oxide fuel assembly with ferritic/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd/KgM and a peak fast fluence greater than 15 x 10 22 n/cm 2 . The cladding, wire-wrap, and duct material for the ACO-1 test assembly is the ferritic/martensitic alloy, HT9, which was chosen for use in long-lifetime fuel assemblies because of its good nominal temperature creep strength and low swelling rate. Valuable experience on the performance of HT9 materials has been gained from this test, advancing our quest for long-lifetime fuel. Pertinent data, obtained from the ACO-1 test assembly, will support the irradiation of the Core Demonstration Experiment in FFTF

FFTF operational results: startup to 100 MWd/kg

International Nuclear Information System (INIS)

Baird, Q.L.; Harris, R.A.

1985-01-01

The Fast Flux Test Facility (FFTF) is a 400-MW(t) sodium-cooled fast reactor operating at the Hanford Engineering Development Laboratory in Richland, Washington, to conduct fuels and materials testing in support of the US liquid-metal fast breeder reactor program. Startup and initial power testing included a comprehensive series of nuclear and nonnuclear tests to verify the thermal and neutronic characteristics of the plant and to demonstrate its inherent safety features. Extensive reactor core characterization measurements were completed to provide the neutron and gamma spectra, fission rates, and other physics data needed to design and evaluate tests irradiated in the FFTF. A specially designed series of natural-circulation tests was performed to demonstrate the inherent safety features of the plant. Early in 1982 the FFTF began its first 100-d irradiation cycle. Since that time the plant has operated beyond expectations; it achieved a cycle capacity factor of 99.5% in the most recent irradiation cycle. One hundred fifty fuel assemblies (80 of which are experiments) and over 32,000 individual fuel pins have been irradiated, some in excess of 100 MWd/kg. Specialized equipment and systems unique to sodium-cooled reactor plants have performed well. There have been no sodium leaks in the 6 y of sodium system operation. Liquid-metal system maintenance techniques have been proven reliable. Plant maintenance and operating personnel radiation exposures have been very low. 15 figs

Seal for turnable lids on nuclear reactors. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Jansing, W; Roehrs, H; Rothfuss, H

1977-08-04

This seal guarantees greatest leak-proofness even when turning the turn-lid and keeps off abrasion and dirt from the interior of the reactor vessel. This is caused by various individual seals in the space above the horizontal flange which closes the vessel at the top. An outer ring is removably supported on an outer diameter of this flange with the narrow side of its polygonal cross-section. An inner ring turns within this body with its narrow side on spheres of a support. The turn-lid is centred with its largest diameter in this inner ring and is supported by a part of the same body protruding it by means of screws. The vertical ring gap between the two ring bodies is sealed by two inflatable hollow seals; they are coated with PTE.

Analysis of ZPPR experiments supporting production of 60Co in FFTF

International Nuclear Information System (INIS)

Wootan, D.W.; Carter, L.L.; Rawlins, J.A.; Schaefer, R.W.; Maddison, D.W.

1987-01-01

An effort to expand the irradiation mission of the Fast Flux Test Facility (FFTF) beyond testing fuels and materials for the liquid-metal reactor (LMR) program included a study of the feasibility of producing commercial quantities of 60 Co. The 60 Co would be produced by neutron capture in assemblies containing an array of natural cobalt pins and hydrogen-bearing moderator pins located at the periphery of the FFTF core. Adding hydrogenous material to the assemblies enhances 60 Co production by slowing neutrons into energy ranges where the 59 Co capture cross section is higher. Some of the moderated neutrons leak from the moderated region to adjacent fuel regions, increasing local fission rates. In order to validate calculated fission rates and establish calculational biases and uncertainties, experiments were conducted in the zero-power plutonium reactor (ZPPR) to measure fission rates in fuel regions adjacent to a mock-up of a cobalt production assembly. The ZPPR experiment and analyses provided the technical basis for irradiating hydride-moderated assemblies in FFTF and a cobalt production cycle of FFTF with generally good agreement between measured and predicted reactivity worth and outlet temperatures in adjacent fuel assemblies. This experiment provides a benchmark configuration of a hydrogenous assembly in a fast reactor, with emphasis on the effect of the hydrogen on adjacent fuel. It is anticipated there will be other applications of hydride moderation in the development of future isotope production missions in FFTF

Leak-rate qualification of the FFTF control area

International Nuclear Information System (INIS)

Billings, M.P.; Swenson, L.D.

1983-06-01

Positive experience with the Fast Flux Test Facility (FFTF) Control Area boundary has demonstrated that strigent requirements for reactor control room leak-tightness can be met and maintained. Guidance contained in 10CFR50, Appendix A, Criteria 4 and 19, and Regulatory Guides 1.78 and 1.95 provided criteria for control room habitability, to provide safe, central control of the FFTF plant under normal and accident conditions. A leak rate criterion of 178 scfm for the approximate 53,000 cu. ft. Volume of the Control Area was established for the limiting condition of airborne sodium oxide aerosols from a postulated fire in one of the three secondary sodium loops. Numerous tests utilizing a variety of leak identification techniques were conducted

Los Alamos Hot-Cell-Facility modifications for examining FFTF fuel pins

International Nuclear Information System (INIS)

Campbell, B.M.; Ledbetter, J.M.

1982-01-01

Commissioned in 1960, the Wing 9 Hot Cell Facility at Los Alamos was recently modified to meet the needs of the 1980s. Because fuel pins from the Fast Flux Test Facility (FFTF) at the Hanford Engineering Development Laboratory (HEDL) are too long for examination in the original hot cells, we modified cells to accommodate longer fuel pins and to provide other capabilities as well. For instance, the T-3 shipping cask now can be opened in an inert atmosphere that can be maintained for all nondestructive and destructive examinations of the fuel pins. The full-length pins are visually examined and photographed, the wire wrap is removed, and fission gas is sampled. After the fuel pin is cropped, a cap is seal-welded on the section containing the fuel column. This section is then transferred to other cells for gamma-scanning, radiography, profilometry, sectioning for metallography, and chemical analysis

1998 FFTF annual system assessment reports

International Nuclear Information System (INIS)

Guttenberg, S.

1998-01-01

The health of FFTF systems was assessed assuming a continued facility standby condition. The review was accomplished in accordance with the guidelines of FFTF-EI-083, Plant Evaluation Program. The attached document includes an executive summary of the significant conclusions and assessment reports for each system evaluated

1998 FFTF annual system assessment reports

Energy Technology Data Exchange (ETDEWEB)

Guttenberg, S.

1998-03-19

The health of FFTF systems was assessed assuming a continued facility standby condition. The review was accomplished in accordance with the guidelines of FFTF-EI-083, Plant Evaluation Program. The attached document includes an executive summary of the significant conclusions and assessment reports for each system evaluated.

CLOSURE OF THE FAST FLUX TEST FACILITY (FFTF) CURRENT STATUS & FUTURE PLANS

Energy Technology Data Exchange (ETDEWEB)

LESPERANCE, C.P.

2007-05-23

The Fast Flux Test Facility (FFTF) was a 400 MWt sodium-cooled fast reactor situated on the U.S. Department of Energy's (DOE) Hanford Site in the southeastern portion of Washington State. DOE issued the final order to shut down the facility in 2001, when it was concluded that there was no longer a need for FFTF. Deactivation activities are in progress to remove or stabilize major hazards and deactivate systems to achieve end points documented in the project baseline. The reactor has been defueled, and approximately 97% of the fuel has been removed from the facility. Approximately 97% of the sodium has been drained from the plant's systems and placed into an on-site Sodium Storage Facility. The residual sodium will be kept frozen under a blanket of inert gas until it is removed later as part of the facility's decontamination and decommissioning (D&D). Plant systems have been shut down and placed in a low-risk state to minimize requirements for surveillance and maintenance. D&D work cannot begin until an Environmental Impact Statement has been prepared to evaluate various end state options and to provide a basis for selecting one of the options. The Environmental Impact Statement is expected to be issued in 2009.

USA/FBR program status FFTF operations startup experience

International Nuclear Information System (INIS)

Moffitt, W.C.; Izatt, R.D.

1981-06-01

This paper gives highlights of the major Operations evaluations and operational results of the startup acceptance testing program and initiation of normal operating cycles for experiment irradiation in the FFTF. 33 figures

FFTF reactor-characterization program: gamma-ray measurements and shield characterization

International Nuclear Information System (INIS)

Bunch, W.L.; Moore, F.S. Jr.

1983-02-01

A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)

FFTF operations procedures preparation guide. Revision 2

International Nuclear Information System (INIS)

1976-12-01

The Guide is intended to provide guidelines for the initial preparation of FFTF Operating Procedures. The Procedures Preparation Guide was developed from the plan presented and approved in the FFTF Reactor Plant Procedures Plan, PC-1, Revision 3

Engineering scale tests of an FFTF fission gas delay bed

International Nuclear Information System (INIS)

Kabele, T.J.; Bohringer, A.P.

1975-01-01

The dynamic adsorption coefficient of 85 Kr on activated charcoal from a nitrogen carrier gas was measured at -80 and -120 0 C at pressures of zero and 30 psig. The effects of the presence of impurities in the nitrogen carrier gas (1 percent oxygen, and 100 vppm carbon dioxide) on the adsorption coefficient of 85 Kr were also measured. The 85 Kr adsorption coefficient increased with decreasing temperature, and increased with increasing pressure. The presence of oxygen and carbon dioxide impurities in the nitrogen carrier gas had no discernible effect upon the adsorption coefficient. The adsorption coefficient for 85 Kr from nitrogen gas was lower than for adsorption of 85 Kr from an argon gas stream. The work concluded a test program which provided design data for the fission gas delay beds which will be installed in the Fast Flux Test Facility (FFTF). (U.S.)

FFTF operations - 1980 - a year of achievement

International Nuclear Information System (INIS)

Newland, D.J.; Bliss, R.J.

1981-01-01

The year 1980 saw FFTF achieve initial criticality, proceed through an intense maintenance and system testing period, and achieve a remarkably successful two-day full power (400 MW) demonstration run. These achievements were preceded by two and one-half years of plant startup activity

TS-1 and TS-2 transient overpower tests on FFTF fuel

International Nuclear Information System (INIS)

Pitner, A.L.; Ferrell, P.C.; Culley, G.E.; Weber, E.T.

1985-01-01

The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P 0 > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences

Strategies in development of advanced fuels for LMFBR

International Nuclear Information System (INIS)

Handa, Muneo

1976-12-01

Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

CLOSURE OF THE FAST FLUX TEST FACILITY (FFTF) CURRENT STATUS and FUTURE PLANS

International Nuclear Information System (INIS)

LESPERANCE, C.P.

2007-01-01

The Fast Flux Test Facility (FFTF) was a 400 MWt sodium-cooled fast reactor situated on the U.S. Department of Energy's (DOE) Hanford Site in the southeastern portion of Washington State. DOE issued the final order to shut down the facility in 2001, when it was concluded that there was no longer a need for FFTF. Deactivation activities are in progress to remove or stabilize major hazards and deactivate systems to achieve end points documented in the project baseline. The reactor has been defueled, and approximately 97% of the fuel has been removed from the facility. Approximately 97% of the sodium has been drained from the plant's systems and placed into an on-site Sodium Storage Facility. The residual sodium will be kept frozen under a blanket of inert gas until it is removed later as part of the facility's decontamination and decommissioning (D and D). Plant systems have been shut down and placed in a low-risk state to minimize requirements for surveillance and maintenance. D and D work cannot begin until an Environmental Impact Statement has been prepared to evaluate various end state options and to provide a basis for selecting one of the options. The Environmental Impact Statement is expected to be issued in 2009

Dissolution of FFTF vendor fuel

International Nuclear Information System (INIS)

Lerch, R.E.

1979-08-01

Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone

Dissolution of FFTF vendor fuel

Energy Technology Data Exchange (ETDEWEB)

Lerch, R.E.

1979-08-01

Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone.

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-09-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

Use of a personal computer for instrumentation testing at the FFTF

International Nuclear Information System (INIS)

Clark, D.A.

1985-01-01

The FFTF uses computers to assist in performance of instrumentation calibration checks for Technical Specification compliance. When the computers used for these checks are not available, back-up methods must be used. A method of using a personal computer to serve as a back-up is discussed, and the hardware used is described

FFTF Work Control Center

International Nuclear Information System (INIS)

Talbot, M.D.

1986-01-01

A centralized Work Control Center (WCC) is responsible for assuring that maintenance and modification of the Fast Flux Test Facility (FFTF) is performed in accordance with written procedures that ensure design integrity, personnel and public safety, and equipment and system availability for the computerized Master Information Data Acquisition System (MIDAS). Each maintenance task is logged into MIDAS from a Work Request from that has been reviewed and prioritized by the WCC. Thereafter, MIDAS is used to track schedule, manpower and material requirements; authorize field work; and close out the maintenance activity

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-01-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

FFTF [Fast Flux Test Facility]/IEM [Interim Examination and Maintenance] Cell Fuel Pin Weighing System

International Nuclear Information System (INIS)

Gibbons, P.W.

1987-09-01

A Fuel Pin Weighing Machine has been developed for use in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell to assist in identifying an individual breached fuel pin from its fuel assembly pin bundle. A weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF) at Hanford, was used as the basis for the IEM Cell system. Design modifications to the original equipment were centered around: 1) adapting the FMEF machine for use in the IEM Cell and 2) correcting operational deficiencies discovered during functional testing in the IEM Cell Mockup

FFTF in-containment cell liner design and installation experience

International Nuclear Information System (INIS)

Umek, A.M.; Swenson, L.D.

1980-01-01

Design features and liner construction techniques are discussed. Cell leak-rate tests and the methods used to locate and repair leaks are described. A brief analysis of the overall experience at FFTF is provided, with recommendations for future plant designs

Design of the core support and restraint structures for FFTF and CRBRP

International Nuclear Information System (INIS)

Sutton, H.G.; Rylatt, J.A.

1977-12-01

This paper presents and compares the design and fabrication of the FFTF and CRBRP reactor structures which support and restrain the reactor core assemblies. The fabrication of the core support structure (CSS) for the FFTF reactor was completed October 1972 and this paper discusses how the fabrication problems encountered with the FFTF were avoided in the subsequent design of the CRBR CSS. The radial core restraint structure of the FFTF was designed and fabricated such that an active system could replace the present passive system which is segmented and relies on the CSS core barrel for total structure integrity to maintain core geometry. The CRBR core restraint structure is designed for passive restraint only, and this paper discusses how the combined strengths of the restraint structure former rings and the CSS core barrel are utilized to maintain core geometry. Whereas the CSS for the FFTF interfaces directly with the reactor core assemblies, the CRBR CSS does not. A comparison is made on how intermediate structures in CRBR (inlet modules) provide the necessary design interfaces for supporting and providing flow distribution to the reactor core assemblies. A discussion is given on how the CRBR CSS satisfied the design requirements of the Equipment Specification, including thermal transient, dynamic and seismic loadings, and results of flow distribution testing that supported the CRBR design effort. The approach taken to simplify fabrication of the CRBR components, and a novel 20 inch deep narrow gap weld joint in the CSS are described

Sodium technology. 1 - FFTF support work: friction tests, January-March 1974

International Nuclear Information System (INIS)

Kurzeka, W.J.; Oliva, R.M.; Horton, P.

1974-01-01

The objective of this program is to conduct friction screening tests in an environment of high-temperature, high-purity liquid sodium or sodium vapor to: (1) develop backup materials, processes, and vendors for core component wear pads, (2) investigate material treatments and coatings for improvement of wear behavior of common LMFBR structural materials, (3) evaluate weld-deposited hardfacings and/or prefabricated bearing materials for use in long-term, high-temperature, high-fluence regions, (4) evaluate bearing materials with a low potential for change in surface composition due to corrosion or mass transfer effects, and (5) develop statistical confidence in friction values for selected material combinations

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

Energy Technology Data Exchange (ETDEWEB)

GANTT, D.A.

2000-10-31

This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

Analytical throughput-estimating methods for the Hot Fuel Examination Facility

International Nuclear Information System (INIS)

Keyes, R.W.; Phipps, R.D.

1983-01-01

The Hot Fuel Examination Facility (HFEF) supports the operation and experimental programs of the major Liquid Metal Fast Breeder Reactor (LMFBR) test facilities; specifically, the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor II (EBR-II), and the Transient Reactor Test (TREAT) Facility. Successful management of HFEF and of LMFBR safety and fuels and materials programs, therefore, requires reliable information regarding HFEF's capability to handle expected or proposed program work loads. This paper describes the 10-step method that has been developed to consider all variables which significantly affect the HFEF examination throughput and quickly provide the necessary planning information

FFTF criteria for run to cladding breach experiments

International Nuclear Information System (INIS)

Van Keuren, J.C.; Heard, F.J.; Stepnewski, D.D.

1985-12-01

The review of experiments proposed for irradiation in FFTF resulted in the development of new criteria for run-to-cladding breach experiments. These criteria have allowed irradiation of aggressive experiments without compromising the safety bases for FFTF. This paper consisting of a set of narrated slides, discusses these criteria and related bases

FFTF Heat Transport System (HTS) component and system design

International Nuclear Information System (INIS)

Young, M.W.; Edwards, P.A.

1980-01-01

The FFTF Heat Transport Systems and Components designs have been completed and successfully tested at isothermal conditions up to 427 0 C (800 0 F). General performance has been as predicted in the design analyses. Operational flexibility and reliability have been outstanding throughout the test program. The components and systems have been demonstrated ready to support reactor powered operation testing planned later in 1980

FFTF/IEM [Fast Flux Test Facility/Interim Examination and Maintenance] cell fuel pin weighing system: Remote maintenance design considerations

International Nuclear Information System (INIS)

Gibbons, P.W.

1986-06-01

A Fuel Pin Weighing Machine has been developed for use in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell to assist in identifying an individual breached fuel pin from its fuel assembly pin bundle. Optimum configuration for remote maintenance was a major consideration in the design of each element of the Pin Weighing System

FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

International Nuclear Information System (INIS)

Stover, R.L.; Beaver, T.R.; Chang, S.C.

1983-01-01

The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

Use of reliability in the LMFBR industry

International Nuclear Information System (INIS)

Penland, J.R.; Smith, A.M.; Goeser, D.K.

1977-01-01

This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

Summary of ORNL long-term surveillance at the FFTF

International Nuclear Information System (INIS)

Damiano, B.; Thie, J.A.

1992-08-01

Oak Ridge National Laboratory has used an automated system between 1983 and 1987 to collect and analyze primary system noise data at the Fast Flux Test Facility (FFTF) located in Hanford, Washington, System operation and data handling are described, data collection efforts are summarized, and principal findings are presented

Load-deflection characteristics of small bore insulated pipe clamps

International Nuclear Information System (INIS)

Severud, L.K.; Clark, G.L.

1982-01-01

High temperature LMFBR piping is subject to rapid temperature changes during transient events. Typically, this pipe is supported by specially designed insulated pipe clamps to prevent excessive thermal stress from developing during these transients. The special insulated clamps used on both FFTF and CRBR piping utilize a Belleville spring arrangement to compensate for pipe thermal expansion. Analysis indicates that this produces a non-linear, directionally sensitive clamp spring rate. Since these spring rates influence the seismic response of a supported piping system, it was deemed necessary to evaluate them further by test. This has been accomplished for the FFTF clamps. A more standard insulated pipe clamp, which does not incorporate Belleville springs to accommodate thermal expansion, was also tested. This type clamp is simple in design, and economically attractive. It may have wide application prospects for use in LMFBR small bore auxiliary piping operating at temperatures below 427 0 C. Load deflection tests were conducted on 2.54 CM and 7.62 CM diameter samples of these commercial clamps

Handling of sodium for the FFTF

International Nuclear Information System (INIS)

Ballif, J.L.; Meadows, G.E.

1978-06-01

Based on the High Temperature Sodium Facility (HTSF) experience and the extensive design efforts for FFTF, procedures are in place for the unloading of the tank cars and for the fill of the FFTF reactor. Special precautions have been taken to provide safe handling and to accommodate contingencies in operation. These contingencies include special protective suits allowing personnel to enter and correct conditions arising from fill operations in the course of moving 7.71 x 10 5 kg (1.7 x 10 6 lbs) of sodium from the tank cars into the reactor vessel and its loop system

Preheat operating experiences at the FFTF

International Nuclear Information System (INIS)

Tucker, W.R.

1978-01-01

The rather extensive test program performed on the FFTF preheat control system resulted in successful sodium fill of one secondary heat transport loop on July 2, 1978. The data obtained during testing and the attendant operating experience gained resulted in some design changes and provided the information necessary to fully characterize system performance. Temperature excursions and deviations from preset limits of only a minor nature were encountered during preheat for sodium fill. The addition of the rate alarm feature was beneficial to operation of the preheat system and allowed early detection and correction of impending excursions

Initial field testing definition of subsurface sealing and backfilling tests in unsaturated tuff

International Nuclear Information System (INIS)

Fernandez, J.A.; Case, J.B.; Tyburski, J.R.

1993-05-01

This report contains an initial definition of the field tests proposed for the Yucca Mountain Project repository sealing program. The tests are intended to resolve various performance and emplacement concerns. Examples of concerns to be addressed include achieving selected hydrologic and structural requirements for seals, removing portions of the shaft liner, excavating keyways, emplacing cementitious and earthen seals, reducing the impact of fines on the hydraulic conductivity of fractures, efficient grouting of fracture zones, sealing of exploratory boreholes, and controlling the flow of water by using engineered designs. Ten discrete tests are proposed to address these and other concerns. These tests are divided into two groups: Seal component tests and performance confirmation tests. The seal component tests are thorough small-scale in situ tests, the intermediate-scale borehole seal tests, the fracture grouting tests, the surface backfill tests, and the grouted rock mass tests. The seal system tests are the seepage control tests, the backfill tests, the bulkhead test in the Calico Hills unit, the large-scale shaft seal and shaft fill tests, and the remote borehole sealing tests. The tests are proposed to be performed in six discrete areas, including welded and non-welded environments, primarily located outside the potential repository area. The final selection of sealing tests will depend on the nature of the geologic and hydrologic conditions encountered during the development of the Exploratory Studies Facility and detailed numerical analyses. Tests are likely to be performed both before and after License Application

Pipe supports and anchors - LMFBR applications

International Nuclear Information System (INIS)

Anderson, M.J.

1983-06-01

Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

FFTF reactor plant procedures plan

International Nuclear Information System (INIS)

1972-01-01

The document presented defines the plan to be used to coordinate the preparation, review, approval, and issuance of the operating procedure documents required to ensure safe and efficient operation of FFTF

Automated refueling inventory control system at FFTF

International Nuclear Information System (INIS)

Ordonez, C.R.

1983-10-01

The Refueling Inventory Control System (RICS) at the Fast Flux Test Facility (FFTF) keeps track of all assemblies that reside in the various refueling facilities. The automated RICS allows the user to obtain information regarding any assembly under its control by displaying the data on a screen. It also provides a simulation mode which allows assembly moves on a duplicated data base. This simulation is used to verify the refueling documentation before it is issued

FFTF disposable solid waste cask

Energy Technology Data Exchange (ETDEWEB)

Thomson, J. D.; Goetsch, S. D.

1983-01-01

Disposal of radioactive waste from the Fast Flux Test Facility (FFTF) will utilize a Disposable Solid Waste Cask (DSWC) for the transport and burial of irradiated stainless steel and inconel materials. Retrievability coupled with the desire for minimal facilities and labor costs at the disposal site identified the need for the DSWC. Design requirements for this system were patterned after Type B packages as outlined in 10 CFR 71 with a few exceptions based on site and payload requirements. A summary of the design basis, supporting analytical methods and fabrication practices developed to deploy the DSWC is provided in this paper.

FFTF disposable solid waste cask

International Nuclear Information System (INIS)

Thomson, J.D.; Goetsch, S.D.

1983-01-01

Disposal of radioactive waste from the Fast Flux Test Facility (FFTF) will utilize a Disposable Solid Waste Cask (DSWC) for the transport and burial of irradiated stainless steel and inconel materials. Retrievability coupled with the desire for minimal facilities and labor costs at the disposal site identified the need for the DSWC. Design requirements for this system were patterned after Type B packages as outlined in 10 CFR 71 with a few exceptions based on site and payload requirements. A summary of the design basis, supporting analytical methods and fabrication practices developed to deploy the DSWC is provided in this paper

Radioactive material package seal tests

International Nuclear Information System (INIS)

Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

1990-01-01

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (US Nuclear Regulatory Commission, 1983). The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. A seal technology program, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management (EM) and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this program to characterize the behavior of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer low-temperature ratings are based on methods that simulate this use. The seal materials tested in this program with a fixture similar to a RAM cask closure, with the exception of silicone S613-60, are not leak tight (1.0 x 10 -7 std cm 3 /s) at manufacturer low-temperature ratings. 8 refs., 3 figs., 1 tab

Performance of FFTF reference fuel and control assemblies

International Nuclear Information System (INIS)

Leggett, R.D.; Weber, E.T.

1984-11-01

This paper describes the performance of the reference fuel and control assemblies used in FFTF through the first four cycles of irradiation (446 equivalent full power days, EFPD). These assemblies performed flawlessly through the rigors of the Startup Testing Program, STP, (beginning in late 1979) with its cyclic operation and continued to do so throughout Cycles 1, 2, 3 and 4, the latter ending in April 1984

FFTF control system experience

International Nuclear Information System (INIS)

Warrick, R.P.

1981-01-01

The FFTF control systems provide control equipment for safe and efficient operation of the plant. For convenience, these systems will be divided into three parts for discussions: (1) Plant Protection System (PPS); (2) Plant Control System (PCS); and (3) General Observations. Performance of each of these systems is discussed

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Breeder reactor program in the USA

International Nuclear Information System (INIS)

Brewer, S.

1978-01-01

In the United States, commercial fuel reprocessing and demonstration test of plutonium breeder reactors were now postponed. LMFBR project and schedule of FFTF and afterwards await the results of INFCE. However, this is not discarding the development of LMFBRs. With the existing energy resources, the United States can have the opportunity to make careful and thorough study. Emphasis is placed on the research and development on new alternative types of fuel. FFTF going to be operated soon should provide effective means for the developments of FBR fuel and materials. High priority is to be retained for the test and development of sodium system hardwares. The nuclear proliferation problem is not related to heat transfer and secondary systems; it is associated with the selection of fuel and fuel cycle. The whole program is centered around LMFBR design and development. The target output will be 600 x 10 3 -- 700 x 10 3 MW. In the United States, now is the time to develop excellent products and to study the nuclear proliferation problem more carefully. (Mori, K.)

Reactor component inventory system at FFTF

International Nuclear Information System (INIS)

Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

1985-02-01

A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

Addendum to ''Test Plan: Small-Scale Seal Performance Tests (SSSPT)''

International Nuclear Information System (INIS)

Finley, R.E.

1992-01-01

This document describes activities that are intended to update the data base of fluid flow measurements made on expansive salt concrete (ESC) seals as part of the Small-Scale Seal Performance Tests (SSSPT). The original plans for the SSSPT experiments are described by Stormont (1985a and 1985b). These seals have previously been tested with brine and gas during the early stages (less than 450 days) after emplacement. The purpose of this Test Plan Addendum is to detail the activities necessary to, update the gas and brine measurements previously performed on the SSSPT Series A and Series B seals and to identify the key personnel responsible for implementing these activities. This addendum describes a limited undertaking and does NOT change the scope of the original test plan

Tamper-indicating devices and safeguards seals evaluation test report

International Nuclear Information System (INIS)

Horton, P.R.V.; Waddoups, I.G.

1993-12-01

Sandia National Laboratories was asked to evaluate the seals used as tamper-indicating devices (TIDs) at DOE facilities. Initially, a survey determined what seal manufacturers were being used and what similar seal types were available. Once the required specifications for TIDs were defined, a test plan measured the currently available seals against the requirements. Environmental and physical type tests stressed the seals under two broad categories: (1) handling durability and (2) tamper resistance. Results of the testing provide comparative ratings for the various seals, recommendations for using currently available seals, and a new tamper-indicating technology

Radiological considerations of the reactor cover gas processing system at the FFTF

International Nuclear Information System (INIS)

Prevo, P.R.

1987-01-01

Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986). (author)

Subcriticality calculations for the FFTF reverse approach to critical experiment

International Nuclear Information System (INIS)

Selby, D.L.; Flanagan, G.F.

1975-01-01

The reverse approach to critical (RAC) experiments were performed in the ZPR-IX critical facility at Argonne National Laboratory. One of the major objectives of this project is to determine the adequacy of the low-level flux monitor (LLFM) detectors for initial loading of the Fast Flux Test Facility (FFTF). 5 references

FFTF sodium and cover gas characterization and purification

International Nuclear Information System (INIS)

McCown, J.J.; Bloom, G.R.; Meadows, G.E.; Mettler, G.W.

1980-02-01

The FFTF Primary and Secondary Heat Transport System (HTS) sodium is purified with cold traps which have packed crystallizers and external economizers. The Primary HTS cold trap is NaK cooled and the Secondary HTS cold traps are air cooled. The FFTF cold traps have maintained high purity in the sodium since sodium fill. Plant operational procedures during fill and initial sodium heatup to 800 0 F were controlled to assure low release rates of impurities to the sodium. The FFTF sodium systems are monitored by plugging temperature indicators and by several sampling methods. During reactor fill and non-fueled operations at 400 to 800 0 F, impurity changes in the sodium were followed by continuous plugging indicator coverage, by exposing wires and foils to measure carbon, hydrogen and oxygen, and by bulk sample analysis of all other trace constituents. The sampling and analysis methods and data are presented, impurity excursions in the cover gas and sodium are described, and impurity trends are discussed

The effect of swelling in Inconel 600 on the performance of FFTF [Fast Flux Test Facility] reflector assemblies

International Nuclear Information System (INIS)

Makenas, B.J.; Trenchard, R.G.; Hecht, S.L.; McCarthy, J.M.; Garner, F.A.

1986-02-01

The Fast Flux Test Facility (FFTF) is designed with non-fueled outer row assemblies, each of which consists of a stack of Inconel 600 blocks penetrated by 316 stainless steel (SS) coolant tubes. These assemblies act as a radial neutron reflector and as a straight but flexible core boundary. During an FFTF refueling outage it was observed that the degree of difficulty in withdrawing an outer row driver fuel assembly was a function of the peak fast fluence of neighboring reflector assemblies. It was subsequently determined through various postirradiation examinations that the reflector assemblies were both bowed and stiff. Measurements of the individual Inconel 600 blocks indicated that the blocks had distorted into a trapezoidal cross section due to differential swelling of Inconel 600 in a steep radial flux gradient. Immersion density results indicate greater irradiation induced volumetric swelling than any previously reported data or correlation for Inconel 600 at equivalent fast fluence. The Inconel 600 swelled approximately the same amount as the SA 316 SS reflector components. Transmission electron microscopy studies on the Inconel blocks and swelling measurements on related materials have been performed and these data have been related to the performance of the reflector materials

CRBRP design and test results for fuel handling systems, plugs, and seals

International Nuclear Information System (INIS)

Berg, G.E.

1977-01-01

The fuel handling system and reactor rotating plugs for the Clinch River Breeder Reactor Plant (CRBRP) are based primarily on existing technology and, in many respects, follow the concept developed for the Fast Flux Test Facility (FFTF). The equipment and the development programs initiated to verify its performance are described. Test results obtained from the development program, and the extent to which these results verified original design selections, or suggested potential improvements, are discussed

LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

Integrated quality status and inventory tracking system for FFTF driver fuel pins

International Nuclear Information System (INIS)

Gottschalk, G.P.

1979-11-01

An integrated system for quality status and inventory tracking of Fast Flux Test Facility (FFTF) driver fuel pins has been developed. Automated fuel pin identification systems, a distributed computer network, and a data base are used to implement the tracking system

DALIS: a computer-assisted document retrieval system for the FFTF

International Nuclear Information System (INIS)

Harves, W.G.

1981-01-01

The FFTF (Fast Flux Test Facility) is a liquid sodium cooled, fast flux reactor designed specifically for irradiation testing of fuels and components for liquid metal fast breeder reactors. The Department of Energy and the Nuclear Regulatory Commission require that all pertinent documentation for maintenance, operation, and safety of the FFTF be readily accessible and retrievable, both during initial startup and for the lifetime of the plant. That amounts to a lot of information which has to be retrievable. The indexing system finally developed is called the DALIS system, short for Document and Location Indexing System. This system was designed by an engineer (Michael Theo) for use by engineers. DALIS uses descriptiors and keywords to identify each document in the system. The descriptors give such information as document number, date of issuance of the document, the title, the originating organization, and the microfilm or hardcopy location of the document. The keywords are words or phrases that describe the content of the document and permit retrieval by means of a computer search for documents with the stated keywords

Large-scale tests of aqueous scrubber systems for LMFBR vented containment

International Nuclear Information System (INIS)

McCormack, J.D.; Hilliard, R.K.; Postma, A.K.

1980-01-01

Six large-scale air cleaning tests performed in the Containment Systems Test Facility (CSTF) are described. The test conditions simulated those postulated for hypothetical accidents in an LMFBR involving containment venting to control hydrogen concentration and containment overpressure. Sodium aerosols were generated by continously spraying sodium into air and adding steam and/or carbon dioxide to create the desired Na 2 O 2 , Na 2 CO 3 or NaOH aerosol. Two air cleaning systems were tested: (a) spray quench chamber, educator venturi scrubber and high efficiency fibrous scrubber in series; and (b) the same except with the spray quench chamber eliminated. The gas flow rates ranged up to 0.8 m 3 /s (1700 acfm) at temperatures to 313 0 C (600 0 F). Quantities of aerosol removed from the gas stream ranged up to 700 kg per test. The systems performed very satisfactorily with overall aerosol mass removal efficiencies exceeding 99.9% in each test

FFTF operational experience

International Nuclear Information System (INIS)

Newland, D.J.; Krupar, J.J.

1984-01-01

In April 1982, the FFTF began its first nominally 100 day irradiation cycle. Since that time the plant has operated very well with steadily increasing plant capacity factors during its first four cycles. One hundred fifty fuel assemblies (eighty of which are experiments) and over 32,000 individual fuel pins have been irradiated, some in excess of 100 MWd/Kg burnup. Specialized equipment and systems unique to sodium cooled reactor plants have performed well

FFTF primary system transition to natural circulation from low reactor power

International Nuclear Information System (INIS)

Bouchey, G.D.; Additon, S.L.; Nutt, W.T.

1980-01-01

Plans for reactor and primary loop natural circulation testing in the Fast Flux Test Facility (FFTF) are summarized. Detailed pretest planning with an emphasis on understanding the implications of process noise and model uncertainties for model verification and test acceptance are discussed for a transition to natural circulation in the reactor core and primary heat transport loops from initial conditions of 5% of rated reactor power and 75% of full flow

UPTF loop seal tests and their RELAP simulation

International Nuclear Information System (INIS)

Tuomainen, M.; Tuunanen, J.

1997-01-01

In a pressurized water reactor the loop seals have an effect on the natural circulation. If a loop seal is filled with water it can cause a flow stagnation in the loop during two-phase natural circulation. Also the pressure loss over a filled loop seal is high, which lowers the water level in the core. Tests to investigate the loop seal behaviour were performed on a German Upper Plenum Test Facility (UPTF). The purpose of the tests was to study the amount of water in the loop seal under different steam flow rates. The tests were simulated with RELAP5/MOD3.2. With high steam flow rates the code had problems in simulating the amount of the water remaining in the pump elbow, but in general the agreement between the calculated results and the experimental data was good. (orig.)

FFTF Plant transition function analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-09-01

The document contains the functions, function definitions, function interfaces, function interface definitions, Input Computer Automated Manufacturing Definition (IDEFO) diagrams, and function hierarchy charts that describe what needs to be performed to deactivate FFTF

Testing of elastomer seals using small-size rigs

International Nuclear Information System (INIS)

Leeks, C.W.E.; Dunford, B.; Barnfield, J.H.; Gray, I.L.S.

1997-01-01

This paper looks at the use of small size seal leakage test rigs to demonstrate the compliance of full size container seals against the IAEA Transport Regulation's limits for activity release for normal transport and accident conditions. The detailed requirements of the regulations are discussed and it is concluded that an appropriate test programme to meet these requirements using only small size test rigs, can normally be set up and carried out on a relatively short time scale. It is important that any small test rigs should be designed to represent the relevant features of the seal arrangement and the overall test programme should cover all of the conditions, specified by the regulations, for the type, classification and contents of the container under consideration. The parameters of elastomer O-rings, which affect their sealing ability, are considered and those which are amenable to small scale testing or have to be modelled at full size are identified. Generally, the seals used in leakage tests have to be modelled with a full size cross-section but can have a reduced peripheral length. (Author)

FFTF fuel failure detection and characterization by cover gas monitoring. Final report

International Nuclear Information System (INIS)

Miller, W.C.; Holt, F.E.

1977-01-01

The Fast Flux Test Facility (FFTF) will include a Fuel Failure Monitoring (FFM) System designed to detect, characterize, and locate fuel and absorber pin failures (i.e., cladding breaches) using a combination of delayed neutron detection, cover gas radioisotope monitoring, and gas tagging. During the past several years the Hanford Engineering Development Laboratory has been involved in the development, design, procurement, and installation of this integrated system. The paper describes one portion of the FFM System, the Cover Gas Monitoring System (CGMS), which has the primary function of fuel failure detection and characterization in the FFTF. By monitoring the various radioisotopes in the cover gas, the CGMS will both detect fuel and absorber pin failures and characterize those failures as to magnitude and severity

Computer simulation of natural circulation in FFTF secondary loops

International Nuclear Information System (INIS)

Beaver, T.R.; Turner, D.M.; Additon, S.L.

1979-07-01

A thermal/hydraulic model of the FFTF secondary heat transport loop has been calibrated against transient natural circulation test data collected March to May 1979. The tests verified that the transition to natural convective flow could be effected from near isothermal conditions without excessive cooling at the air dump heat exchangers. Key empirical parameters of pressure drop and heat loss were found to be at 88% and 81% of pretest estimates, respectively. Pretest piping thermal transport and flow calculational models required no further revision to produce good agreement with test data

Fabrication of FFTF fuel pin wire wrap

International Nuclear Information System (INIS)

Epperson, E.M.

1980-06-01

Lateral spacing between FFTF fuel pins is required to provide a passageway for the sodium coolant to flow over each pin to remove heat generated by the fission process. This spacing is provided by wrapping each fuel pin with type 316 stainless steel wire. This wire has a 1.435mm (0.0565 in.) to 1.448mm (0.0570 in.) diameter, contains 17 +- 2% cold work and was fabricated and tested to exacting RDT Standards. About 500 kg (1100 lbs) or 39 Km (24 miles) of fuel pin wrap wire is used in each core loading. Fabrication procedures and quality assurance tests are described

MWTF jumper connector integral seal block development and leak testing

International Nuclear Information System (INIS)

Ruff, E.S.; Jordan, S.R.

1995-01-01

In fiscal year 1993, tests of an o-ring/tetraseal retainer designed to replace a gasket-type seal used in PUREX-type process jumper connectors encouraged the design of an improved seal block. This new seal block combines several parts into one unitized component called an integral seal block. This report summarizes development and leak testing of the new integral seal block. The integral seal block uses a standard o-ring nested in a groove to accomplish leak tightness. This seal block eliminates the need to machine acme threads into the lower skirt casting and seal retainers, eliminates tolerance stack-up, reduces parts inventory, and eliminates an unnecessary leak path in the jumper connector assembly. This report also includes test data on various types of o-ring materials subjected to heat and pressure. Materials tested included Viton, Kalrez, and fluorosilicone, with some incidental data on teflon coated silicone o-rings. Test experience clearly demonstrates the need to test each seal material for temperature and pressure in its intended application. Some materials advertised as being open-quotes betterclose quotes at higher temperatures did not perform up to expectations. Inspection of the fluorosilicone and Kalrez seals after thermal testing indicates that they are much more susceptible to heat softening than Viton

Status and program of development of the fast breeder reactor system in the U.S

International Nuclear Information System (INIS)

Roberts, R.W.

1977-01-01

The U.S. Energy Research and Development Administration's highest priority breeder concept is the Liquid Metal Fast Breeder Reactor (LMFBR). Its objective is the development of a broad technological and engineering base with extensive utility and industrial involvement which will establish a timely capability for a competitive and commercial breeder industry. In addition to technological development, an integral part of the LMFBR program is the transfer of LMFBR technology to the nuclear industry and the parallel development of utility capability. An important indicator of the effectiveness of this program is the successful construction and operation of a number of LMFBR experimental and test reactors and associated non-nuclear test facilities such as those located at Santa Susana, California, the Argonne (III). National Laboratory and the Hanford Engineering Development Laboratory at Richland, Washington. A principal element of the U.S. fast breeder reactor program is the Fast Flux Test Facility (FFTF). The largest test reactor facility in the world, the FFTF will focus on the testing of fuels and material. The FFTF, being built at the Hanford Engineering Development Laboratory, is scheduled for completion in the late 1970's. The next step toward timely commercial breeder capability is to implement a program for a large-scale demonstration of this concept. A 380 megawatt electrical demonstration plant - designated the Clinch River Breeder Reactor Plant (CRBRP) - will be constructed and operated near Oak Ridge, Tennessee, under a cooperative arrangement of industrial contractors, utilities, and the Government. The completion date is 1983. The estimated cost of the CRBRP project is 1.95 billion dollars. This includes the cost of design, construction, related research and development, and five years of operation - the full timespan from the project's beginning in 1972 through completion. The Nation's principal electric utilities have strongly endorsed the

An automated tensile machine for small specimens heavily neutron irradiated in FFTF/MOTA

International Nuclear Information System (INIS)

Kohyama, Akira; Sato, Shinji; Hamada, Kenichi

1993-01-01

The objective of this work is to develop a fully automated tensile machine for post-irradiation examination (PIE) of Fast Flux Test Facility (FFTF)/Materials Open Test Assembly (MOTA) irradiated miniature tension specimens. The anticipated merit of the automated tensile machine is to reduce damage to specimens during specimen handling for PIE and to reduce exposure to radioactive specimens. This machine is designed for testing at elevated temperatures, up to 873 K, in a vacuum or in an inert gas environment. Twelve specimen assemblies are placed in the vacuum chamber that can be tested successively in a fully automated manner. A unique automated tensile machine for the PIE of FFTF/MOTA irradiated specimens, the Monbusho Automated Tensile Machine (MATRON) consists of a test frame with controlling units and an automated specimen-loading apparatus. The qualification of the test frame has been completed, and the results have satisfied the machine specifications. The capabilities of producing creep and relaxation data have been demonstrated for Cu, Al, 316SS, and ferritic steels. The specimen holders for the three-point bending test and the small bulge test (small punch test; SP test) were also designed and produced

Mechanical response of FFTF reference and P1 cladding tubes under transient heating

International Nuclear Information System (INIS)

Youngahl, C.A.; Ariman, T.; Lepacek, B.E.

1977-01-01

Burst tests of Type 316 stainless steel cladding tube samples subjected to increasing temperature and relatively constant internal pressure were conducted to assist in the pretest analysis of the P1 experiment performed in the Sodium Loop Safety Facility. This paper reports and analyzes the burst test results and those of subsequent transient heating work. The use of a modified extensometer in obtaining mechanical response data for stainless steel in the high temperature range is illustrated, some of such data is provided, and the potential of further experiments and analysis is indicated. Tubing of the same design as Fast Flux Test Facility (FFTF) cladding (20% cold worked, 0.230 in. OD, 15 mil wall) was tested as-received and after annealing or electrolytic thinning. P1 tubing (38% cold worked, 0.230 in. OD, 10 mil wall) was tested before and after aging under conditions anticipated in the P1 reactor experiment. The P1 cladding was designed to simulate FFTF tubing that had experienced irradiation embrittlement and attack by cesium oxide and sodium impurities

Fuel assembly cooling experience at the FFTF/IEM cell

International Nuclear Information System (INIS)

McGuinness, P.W.

1985-01-01

In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs

Monthly report No. 9 (September, October, and November 1977) of INTERATOM representative to Westinghouse Hanford Co., Richland, Washington (USA)

Energy Technology Data Exchange (ETDEWEB)

Mayer, H.

1977-11-30

The following topics are mentioned: alternate breeder fuel cycles (Shippingport Light Water Breeder Reactor), Clinch River Breeder Reactor status, FFTF status, LMFBR component development, and PLBR status. (DG)

Summary of FY 1997 work related to JAPC-U.S. DOE contract study on improvement of core safety - study on GEM (III)

International Nuclear Information System (INIS)

Burke, T.M.

1998-01-01

FFTF was originally designed/constructed/operated to develop LMFBR fuels and materials. Inherent safety became a major focus of the US nuclear industry in the mid 1980's. The inherent safety characteristics of LMFBRs were recognized but additional enhancement was desired. The presentation contents are: Fast Flux Test Facility history and status; Overview of contract activities; Summary of loss of flow without scram with GEMs testing; and Summary of pump start with GEMs testing

Aspects related to the testing of sealed radioactive sources

International Nuclear Information System (INIS)

Olteanu, C. M.; Nistor, V.; Valeca, S. C.

2016-01-01

Sealed radioactive sources are commonly used in a wide range of applications, such as: medical, industrial, agricultural and scientific research. The radioactive material is contained within the sealed source and the device allows the radiation to be used in a controlled way. Accidents can result if the control over a small fraction of those sources is lost. Sealed nuclear sources fall under the category of special form radioactive material, therefore they must meet safety requirements during transport according to regulations. Testing sealed radioactive sources is an important step in the conformity assessment process in order to obtain the design approval. In ICN Pitesti, the Reliability and Testing Laboratory is notified by CNCAN to perform tests on sealed radioactive sources. This paper wants to present aspects of the verifying tests on sealed capsules for Iridium-192 sources in order to demonstrate the compliance with the regulatory requirements and the program of quality assurance of the tests performed. (authors)

Fracture toughness of ferritic alloys irradiated at FFTF

International Nuclear Information System (INIS)

Huang, F.H.

1986-05-01

Ferritic compact tension specimens loaded in the Material Open Test Assembly (MOTA) for irradiation during FFTF Cycle 4 were tested at temperatures ranging from room temperature to 428/degree/C. The electrical potential single specimen method was used to measure the fracture toughness of the specimens. Results showed that the fracture toughness of both HT-9 and 9Cr-1Mo decreases with increasing test temperature and that the toughness of HT-9 was about 30% higher than that of 9Cr-1Mo. In addition, increasing irradiation temperature resulted in an increase in tearing modulus for both alloys. 4 refs., 5 figs., 1 tab

Subscale hood seal test topical report

Energy Technology Data Exchange (ETDEWEB)

Versteeg, J.L.; Herold, B.A.; McClintic, J.K.; Schmall, R.A.; Hoetzl, M.

1991-09-06

To maximize the transfer of heat from the recirculated gases to the scrap, it is essential to percolate as much of the gases as possible through the scrap. To accomplish this flow path and avoid the bypassing of hot gas around the scrap, the seal between the preheater hood and the scrap bucket must be relatively tight. These tests which are described in this report were designed to measure the performance of several possible seal designs under simulated operating conditions. At the conclusion of the tests, one design was recommended as the primary arrangement with another design considered as an alternate. Both designs met the criteria of low leakage but one design was preferred due an expected greater resistance to wear. The test results also provided valuable information for estimating seal leakage in the full scale installation.

Measurement of heat and momentum eddy diffusivities in recirculating LMFBR outlet plenum flows

International Nuclear Information System (INIS)

Manno, V.P.; Golay, M.W.

1978-06-01

An optical technique has been developed for the measurement of the eddy diffusivity of heat in a transparent flowing medium. The method uses a combination of two established measurement tools: a Mach-Zehnder interferometer for the monitoring of turbulently fluctuating temperature and a Laser Doppler Anemometer (LDA) for the measurement of turbulent velocity fluctuations. The technique is applied to the investigation of flow fields characteristic of the LMFBR outlet plenum. The study is accomplished using air as the working fluid in a small scale Plexiglas test section. Lows are introduced into both the 1 / 15 scale FFTF outlet plenum and the 3 / 80 scale CRBR geometry plenum at inlet Reynolds numbers of 22,000. Measurements of the eddy diffusivity of heat and the eddy diffusivity of momentum are performed at a total of 11 measurement stations. Significant differences of the turbulence parameters are found between the two geometries, and the higher chimney structure of the CRBR case is found to be the major cause of the distinction. Spectral intensity studies of the fluctuating electronic analog signals of velocity and temperature are also performed. Error analysis of the overall technique indicates an experimental error of 10% in the determination of the eddy diffusivity of heat and 6% in the evaluation of turbulent momentum viscosity. In general it is seen that the turbulence in the cases observed is not isotropic, and use of isotropic turbulent heat and momentum diffusivities in transport modelling would not be a valid procedure

Feasibility study on large pool-type LMFBR

International Nuclear Information System (INIS)

1984-01-01

A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

Ultrasonic inspection development at HEDL

International Nuclear Information System (INIS)

Day, C.K.; Mech, S.J.; Michaels, T.E.; Dixon, N.E.

1978-01-01

Ultrasonic testing methods and equipment are being developed to support preservice and in-service inspection of selected FFTF welds. A digital computer system is employed in the analysis of both simulated FFTF pipe sections and plate specimens containing fatigue cracks. It is anticipated that test evaluation standards containing fatigue cracks will partially eliminate questions formerly associated with weld test calibration producers by providing natural cracks which follow grain boundaries and stress patterns resembling piping situ conditions. Studies have revealed that commercial transducers may satisfy LMFBR ultrasonic pipe inspection applications: The test system evaluation included transducers and wedge coupling and fluid coupling materials which exhibited acceptable performance at temperatures to 2300C. Results are presented that demonstrate the feasibility of ultrasonic inspection of components immersed in sodium at temperatures to 2600C. (UK)

FFTF railroad tank car Safety Evaluation for Packaging

International Nuclear Information System (INIS)

Carlstrom, R.F.

1995-01-01

This Safety Evaluation for Packaging (SEP) provides evaluations considered necessary to approve transfer of the 8,000 gallon Liquid Waste Tank Car (LWTC) from Fast Flux Test Facility (FFTF) to the 200 Areas. This SEP will demonstrate that the transfer of the LWTC will provide an equivalent degree of safety as would be provided by packages meeting U.S. Department of Transportation (DOT) requirements. This fulfills onsite transportation requirements implemented in the Hazardous Material Packaging and Shipping, WHC-CM-2-14

FFTF metal fuel pin sodium bond quality verification

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1988-12-01

The Fast Flux Test Facility (FFTF) Series III driver fuel design consists of U-10Zr fuel slugs contained in a ferritic alloy cladding. A liquid metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. This document discusses this verification

Breached fuel location in FFTF by delayed neutron monitor triangulation

International Nuclear Information System (INIS)

Bunch, W.L.; Tang, E.L.

1985-10-01

The Fast Flux Test Facility (FFTF) features a three-loop, sodium-cooled 400 MWt mixed oxide fueled reactor designed for the irradiation testing of fuels and materials for use in liquid metal cooled fast reactors. To establish the ultimate capability of a particular fuel design and thereby generate information that will lead to improvements, many of the fuel irradiations are continued until a loss of cladding integrity (failure) occurs. When the cladding fails, fission gas escapes from the fuel pin and enters the reactor cover gas system. If the cladding failure permits the primary sodium to come in contact with the fuel, recoil fission products can enter the sodium. The presence of recoil fission products in the sodium can be detected by monitoring for the presence of delayed neutrons in the coolant. It is the present philosophy to not operate FFTF when a failure has occurred that permits fission fragments to enter the sodium. Thus, it is important that the identity and location of the fuel assembly that contains the failed cladding be established in order that it might be removed from the core. This report discusses method of location of fuel element when cladding is breached

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Test of Seal System for Flexible Pipe End Fitting

DEFF Research Database (Denmark)

Banke, Lars; Jensen, Thomas Gregers

1999-01-01

The purpose of the end fitting seal system is to ensure leak proof termination of flexible pipes. The seal system of an NKT end fitting normally consists of a number of ring joint gaskets mounted in a steel sleeve on the outside of the polymeric inner liner of the pipe. The seal system is activated...... by compression of the gaskets, thus using the geometry to establish a seal towards the inner liner of the pipe and the steel sleeve of the end fitting. This paper describes how the seal system of an end fitting can be tested using an autoclave. By regulating temperature and pressure, the seal system can...... be tested up to 130oC and 51.7 MPa. Pressure, temperature and the mechanical behaviours of the pipe are measured for use in further research. The set-up is used to test the efficiency of the seal system as function of parameters such as cross sectional shapes of the gaskets, tolerances between gaskets...

FFTF fuel pin design procedure verification for transient operation

International Nuclear Information System (INIS)

Baars, R.E.

1975-05-01

The FFTF design procedures for evaluating fuel pin transient performance are briefly reviewed, and data where available are compared with design procedure predictions. Specifically, burst conditions derived from Fuel Cladding Transient Tester (FCTT) tests and from ANL loss-of-flow tests are compared with burst pressures computed using the design procedure upon which the cladding integrity limit was based. Failure times are predicted using the design procedure for evaluation of rapid reactivity insertion accidents, for five unterminated TREAT experiments in which well characterized fuel failures were deliberately incurred. (U.S.)

Development and application of test apparatus for classification of sealed source

International Nuclear Information System (INIS)

Kim, Dong Hak; Seo, Ki Seog; Bang, Kyoung Sik; Lee, Ju Chan; Son, Kwang Je

2007-01-01

Sealed sources have to conducted the tests be done according to the classification requirements for their typical usages in accordance with the relevant domestic notice standard and ISO 2919. After each test, the source shall be examined visually for loss of integrity and pass an appropriate leakage test. Tests to class a sealed source are temperature, external pressure, impact, vibration and puncture test. The environmental test conditions for tests with class numbers are arranged in increasing order of severity. In this study, the apparatus of tests, except the vibration test, were developed and applied to three kinds of sealed source. The conditions of the tests to class a sealed source were stated and the difference between the domestic notice standard and ISO 2919 were considered. And apparatus of the tests were made. Using developed apparatus we conducted the test for 192 Ir brachytherapy sealed source and two kinds of sealed source for industrial radiography. 192 Ir brachytherapy sealed source is classified by temperature class 5, external pressure class 3, impact class 2 and vibration and puncture class 1. Two kinds of sealed source for industrial radiography are classified by temperature class 4, external pressure class 2, impact and puncture class 5 and vibration class 1. After the tests, Liquid nitrogen bubble test and vacuum bubble test were done to evaluate the safety of the sealed sources

FFTF fission gas monitor computer system

International Nuclear Information System (INIS)

Hubbard, J.A.

1987-01-01

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The fission gas monitor system makes extensive use of commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows one monitor to be taken out of service for periodic tests or maintenance without interrupting the overall system functions. A built-in calibrated gamma source can be controlled by the computer, allowing the system to provide rapid system self tests and operational performance reports

SSME Seal Test Program: Test results for sawtooth pattern damper seal

Science.gov (United States)

Childs, D. W.

1986-01-01

Direct and transverse force coefficients for 11, sawtooth-pattern, and damper-seal configurations were examined. The designation damper seal uses a deliberately roughened stator and smooth rotor to increase the net damping force developed by a seal. The designation sawtooth-pattern refers to a stator roughness pattern. The sawtooth pattern yields axial grooves in the stator which are interrupted by spacer elements which act as flow constrictions or dams. All seals use the same smooth rotor and have the same, constant, minimum clearance. The stators examined the consequences of changes in the following design parameters: (1) axial-groove depth; (2) number of teeth: (3) number of sawtooth sections; (4) number of spacer elements; (5) dam width; (6) axially aligned sawtooth sections versus axially-staggered sawtooth sections; and (7) groove geometry. It is found that none of the sawtooth-pattern seal performs as well as the best round-hole-pattern seal. Maximum damping configurations for the sawtooth and round-hole-pattern stators have comparable stiffness performance. Several of the sawtooth pattern stators outperformed the best round-hole pattern seal.

Load-deflection characteristics of small-bore insulated-pipe clamps

International Nuclear Information System (INIS)

Severud, L.K.; Clark, G.L.

1981-12-01

The special insulated clamps used on both FFTF and CRBR piping utilize a Belleville spring arrangement to compensate for pipe thermal expansion. Analysis indicates that this produces a non-linear, directionally sensitive clamp spring rate. Since these spring rates influence the seismic response of a supported piping system, it was deemed necessary to evaluate them further by test. This has been accomplished for the FFTF clamps. A more standard insulated pipe clamp, which does not incorporate Belleville springs to accommodate thermal expansion, was also tested. This type clamp is simple in design, and economically attractive. It may have wide application prospects for use in LMFBR small bore auxiliary piping operating at temperatures below 427 0 C. Load deflection tests were conducted on 2.54 CM and 7.62 CM diameter samples of these commercial clamps

Metal and elastomer seal tests for accelerator applications

International Nuclear Information System (INIS)

Welch, K.M.; McIntyre, G.T.; Tuozzolo, J.E.; Skelton, R.; Pate, D.J.; Gill, S.M.

1989-01-01

The vacuum system of the Alternating Gradient Synchrotron (AGS) at Brookhaven National Laboratory has more than a thousand metal vacuum seals. Also, numerous elastomer seals are used throughout the AGS to seal large beam component chambers. An accelerator upgrade program is being implemented to reduce the AGS operating pressure by x100 and improve the reliability of the vacuum system. This paper describes work in progress on metal and elastomer vacuum seals to help meet those two objectives. Tests are reported on the sealing properties of a variety of metal seals used on different sealing surfaces. Results are also given on reversible sorption properties of certain elastomers. 16 refs., 6 figs., 4 tabs

Reirradiation in FFTF of swelling-resistant Path A alloys previously irradiated in HFIR

International Nuclear Information System (INIS)

Maziasz, P.J.

1985-01-01

Disks of Path A Prime Candidate Alloys (in several pretreatment conditions) and several heats of cold-worked (CW) type 316 and D9 type austenitic stainless steels have been irradiated in HFIR at 300, 500, and 600 0 C to fluences producing about 10 to 44 dpa and 450 to 3600 at. ppm He. These samples are being reirradiated in the Materials Open Test Assembly (MOTA) in FFTF at 500 and 600 0 C, together (side by side) with previously unirradiated disks of exactly the same materials, to greater than 100 dpa. These samples many of which have either very fine helium cluster or helium bubble distributions after HFIR irradiation, are intended to test the possibility and magnitude of a helium-induced extension of the initial low-swelling transient regime relative to the void swelling behavior normally found during FFTF irradiation. Further, these samples will reveal the microstructural stability or evolution differences that correlate with such helium effects. 17 references, 4 tables

Issues in the selection of the LMFBR steam cycle

International Nuclear Information System (INIS)

Buschman, H.W.; McConnell, R.J.

1983-01-01

Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

Hermetic Seal Leak Detection Apparatus with Variable Size Test Chamber

Science.gov (United States)

Kelley, Anthony R. (Inventor)

2015-01-01

The present invention is a versatile hermetic seal leak detection apparatus for testing hermetically sealed containers and devices for leaks without the need to create a custom or specially manufactured testing chamber conforming to the dimensions of the specific object under test. The size of the testing chamber may be mechanically adjusted by the novel use of bellows to reduce and optimize the amount of gas space in a test chamber which surrounds the hermetically sealed object under test. The present invention allows the size of the test chamber to be selectively adjusted during testing to provide an optimum test chamber gas space. The present invention may be further adapted to isolate and test specific portions of the hermetically sealed object under test for leaks.

Small sodium valve design and operating experience

International Nuclear Information System (INIS)

McGough, C.B.

1974-01-01

The United States Liquid Metal Fast Breeder Reactor program (LMFBR) includes an extensive program devoted to the development of small sodium valves. This program is now focused on the development and production of valves for the Fast Flux Test Facility (FFTF) now under construction near Richland, Washington. Other AEC support facilities, such as various test loops located at the Liquid Metal Engineering Center (LMEC), Los Angeles, California, and at the Hanford Engineering Development Laboratory (HEDL), Richland, Washington, also have significant requirements for small sodium valves, and valves similar in design to the FFTF valves are being supplied to these AEC laboratories for use in their critical test installations. A principal motivation for these valve programs, beyond the immediate need to provide high-reliability valves for FFTF and the support facilities, is the necessity to develop small valve technology for the Clinch River Breeder Reactor Plant (CRBRP). FFTF small sodium valve design and development experience will be directly applied to the CRBRP program. Various test programs have been, and are being, conducted to verify the performance and integrity of the FFTF valves, and to uncover any potential problems so that they can be corrected before the valves are placed in service in FFTF. The principal small sodium valve designs being utilized in current U.S. programs, the test and operational experience obtained to date on them, problems uncovered, and future development and testing efforts being planned are reviewed. The standards and requirements to which the valves are being designed and fabricated, the valve designs in current use, valve operators, test and operating experience, and future valve development plans are summarized. (U.S.)

Irradiation of copper alloys in FFTF

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1984-01-01

Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

Testing of neutron-irradiated ceramic-to-metal seals

International Nuclear Information System (INIS)

Brown, R.D.; Clinard, F.W. Jr.; Lopez, M.R.; Martinez, H.; Romero, T.J.; Cook, J.H.; Barr, H.N.; Hittman, F.

1990-01-01

This paper reports on ceramic-to-metal seals prepared by sputtering a titanium metallizing layer onto ceramic disks and then brazing to metal tubes. The ceramics used were alumina, MACOR, spinel, AlON, and a mixture of Al 2 O 3 and Si 3 N 4 . Except for the MACOR, which was brazed to a titanium tube, the ceramics were brazed to niobium tubes. The seals were leak tested and then sent to Los Alamos National Laboratory, where they were irradiated using the spallation neutron source at the Los Alamos Meson Physics Facility. Following irradiation for ∼ 90 days to a fluence of 2.8 x 10 23 n/m 2 , the samples were moved to hot cells and again leak tested. Only the MACOR samples showed any measurable leaks. One set of samples was then pressurized to 6.9 MPa (1000 psi) and subsequently leak tested. No leaks were found. Bursting the seals required hydrostatic pressures of at least 34 MPa (5000 psi). The high seal strength and few leaks indicate that ceramic-to-metal seals can resist radiation-induced degradation

Calculational tracking of decay heat for FFTF plant

International Nuclear Information System (INIS)

Cillan, T.F.; Carter, L.L.

1985-01-01

A detailed calculational monitoring of decay heat for each assembly on the Fast Flux Test Facility (FFTF) plant is obtained by utilizing a decay heat data base and user friendly computer programs to access the data base. Output includes the time-dependent decay heat for an assembly or a specific set of assemblies, and optional information regarding the curies of activated nuclides along the axial length of the assembly. The decay heat data base is updated periodically, usually at the end of each irradiation cycle. 1 ref., 2 figs

Measurement and control of radioactive material transport in FFTF

International Nuclear Information System (INIS)

Maffei, H.P.; Brehm, W.F.; Moore, F.S.; Stinson, W.P.; Bunch, W.L.; Anantatmula, R.P.; McGuire, J.C.

1984-01-01

Techniques and equipment have been developed for measuring radionuclide buildup in FFTF. To date, only 54 Mn and 22 Na have been found in significant quantities. Radiation levels of up to 115 mR/h from 54 Mn and 121 mR/h from 22 Na have been measured in the cells, in reasonable agreement with predicted values. No fission products have been found, since no significant operation with breached fuel has yet been experienced. The buildup of corrosion product and fission product radionuclides is not expected to cause severe maintenance problems in FFTF; should problems arise, corrective measures are being developed

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

A seal analyzer for testing container integrity

International Nuclear Information System (INIS)

McDaniel, P.; Jenkins, C.

1988-01-01

This paper reports on the development of laboratory and production seal analyzer that offers a rapid, nondestructive method of assuring the seal integrity of virtually any type of single or double sealed container. The system can test a broad range of metal cans, drums and trays, membrane-lidded vessels, flexible pouches, aerosol containers, and glass or metal containers with twist-top lids that are used in the chemical/pesticide (hazardous materials/waste), beverage, food, medical and pharmaceutical industries

Thermal expansion movements of piping during FFTF plant startup

International Nuclear Information System (INIS)

Lindquist, M.R.

1981-03-01

FFTF liquid metal piping exhibits significant displacements during heatup of the plant heat transport system. Verification of correct piping movements is important to assure that no restraints are present and to provide data for additional piping design/analysis validation. A test program is described in which a series of measurements were taken at selected piping locations. These data were obtained during Plant Acceptance Testing involving system heatup cycles to approximately 800 0 F(427 0 C). Typical test data are shown and compared to analytical predictions. Two piping system problems that were identified as a result of the testing are described along with resolutions thereof. Establishment of final baseline data is discussed

LMFBR operational safety: the EBR-II experience

International Nuclear Information System (INIS)

Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

1978-01-01

The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

Rock sealing - large scale field test and accessory investigations

International Nuclear Information System (INIS)

Pusch, R.

1988-03-01

The experience from the pilot field test and the basic knowledge extracted from the lab experiments have formed the basis of the planning of a Large Scale Field Test. The intention is to find out how the 'instrument of rock sealing' can be applied to a number of practical cases, where cutting-off and redirection of groundwater flow in repositories are called for. Five field subtests, which are integrated mutually or with other Stripa projects (3D), are proposed. One of them concerns 'near-field' sealing, i.e. sealing of tunnel floors hosting deposition holes, while two involve sealing of 'disturbed' rock around tunnels. The fourth concerns sealing of a natural fracture zone in the 3D area, and this latter test has the expected spin-off effect of obtaining additional information on the general flow pattern around the northeastern wing of the 3D cross. The fifth test is an option of sealing structures in the Validation Drift. The longevity of major grout types is focussed on as the most important part of the 'Accessory Investigations', and detailed plans have been worked out for that purpose. It is foreseen that the continuation of the project, as outlined in this report, will yield suitable methods and grouts for effective and long-lasting sealing of rock for use at stategic points in repositories. (author)

10 CFR 39.35 - Leak testing of sealed sources.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Leak testing of sealed sources. 39.35 Section 39.35 Energy....35 Leak testing of sealed sources. (a) Testing and recordkeeping requirements. Each licensee who uses... record of leak test results in units of microcuries and retain the record for inspection by the...

Initial field testing definition of subsurface sealing and backfilling tests in unsaturated tuff; Yucca Mountain Site Characterization Project

Energy Technology Data Exchange (ETDEWEB)

Fernandez, J.A. [Sandia National Labs., Albuquerque, NM (United States); Case, J.B.; Tyburski, J.R. [I. T. Corp., Albuquerque, NM (United States)

1993-05-01

This report contains an initial definition of the field tests proposed for the Yucca Mountain Project repository sealing program. The tests are intended to resolve various performance and emplacement concerns. Examples of concerns to be addressed include achieving selected hydrologic and structural requirements for seals, removing portions of the shaft liner, excavating keyways, emplacing cementitious and earthen seals, reducing the impact of fines on the hydraulic conductivity of fractures, efficient grouting of fracture zones, sealing of exploratory boreholes, and controlling the flow of water by using engineered designs. Ten discrete tests are proposed to address these and other concerns. These tests are divided into two groups: Seal component tests and performance confirmation tests. The seal component tests are thorough small-scale in situ tests, the intermediate-scale borehole seal tests, the fracture grouting tests, the surface backfill tests, and the grouted rock mass tests. The seal system tests are the seepage control tests, the backfill tests, the bulkhead test in the Calico Hills unit, the large-scale shaft seal and shaft fill tests, and the remote borehole sealing tests. The tests are proposed to be performed in six discrete areas, including welded and non-welded environments, primarily located outside the potential repository area. The final selection of sealing tests will depend on the nature of the geologic and hydrologic conditions encountered during the development of the Exploratory Studies Facility and detailed numerical analyses. Tests are likely to be performed both before and after License Application.

Fluid flow measurements of Test Series A and B for the Small Scale Seal Performance Tests

International Nuclear Information System (INIS)

Peterson, E.W.; Lagus, P.L.; Lie, K.

1987-12-01

The degree of waste isolation achieved by a repository seal system is dependent upon the fluid flow characteristics, or permeability, of the seals. In order to obtain meaningful, site-specific data on the performance of various possible seal system components, a series of in situ experiments called the Small Scale Seal Performance Tests (SSSPT) are being conducted at the Waste Isolation Pilot Plant (WIPP). This report contains the results of gas flow, tracer penetration, and brine flow tests conducted on concrete seals in vertical (Test Series A) and horizontal (Test Series B) configurations. The test objectives were to evaluate the seal performance and to determine if there existed scaling effects which could influence future SSSPT designs. 3 refs., 77 figs

Applicability of the Reactor Safety Study (WASH-1400) to LMFBR risk assessments

International Nuclear Information System (INIS)

El-Sheikh, K.A.; Feller, K.G.; Fleischer, L.; Greebler, P.; McDonald, A.; Sultan, P.; Temme, M.I.; Fullwood, R.R.

1976-01-01

The feasibility of applying the WASH-1400 methods and data to LMFBR risk assessment is evaluated using the following approach for a selected LMFBR: (1) Structuring the LMFBR risk assessment problem in a modular form similar to WASH-1400; (2) Comparing the predictive tools applicable to each module; (3) Comparing the dependencies among the various modules. It is concluded that the WASH-1400 applicability is limited due to LWR-LMFBR differences in operating environments and accident phenomena. WASH-1400 and LMFBR specific methods applicable to LMFBR risk assessments are indicated

Neutron and gamma characterization within the FFTF reactor cavity

International Nuclear Information System (INIS)

Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

1980-08-01

Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

Loss-of-flow test L5 on FFTF-type irradiated fuel

International Nuclear Information System (INIS)

Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

1978-03-01

Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

NEUTRON-INDUCED SWELLING OF Fe-Cr BINARY ALLOYS IN FFTF AT ∼400 DEGREES C

International Nuclear Information System (INIS)

Garner, Francis A.; Greenwood, Lawrence R.; Okita, Taira; Sekimura, Naoto; Wolfer, W. G.

2002-01-01

The purpose of this effort is to determine the influence of dpa rate, He/dpa ratio and composition on the void swelling of simple binary Fe-Cr alloys. Contrary to the behavior of swelling of model fcc Fe-Cr-Ni alloys irradiated in the same FFTF-MOTA experiment, model bcc Fe-Cr alloys do not exhibit a dependence of swelling on dpa rate at approximately 400 degrees C. This is surprising in that an apparent flux-sensitivity was observed in an earlier comparative irradiation of Fe-Cr binaries conducted in EBR-II and FFTF. The difference in behavior is ascribed to the higher helium generation rates of Fe-Cr alloys in EBR-II compared to that of FFTF, and also the fact that lower dpa rates in FFTF are accompanied by progressively lower helium generation rates.

Leak-thight seals got high pressure testing of pipes, tanks, valves

International Nuclear Information System (INIS)

Estrade, J.

1985-01-01

Leak-tight seals ensure quick, safe and efficient testing of pipes with plain-ended or flanged openings, valves with flanged or welded edges, manifields, recipients, etc. They are inserted into the pipe end manually then simply a slight turn of the seal treated wheel commences the pressure test. Hydraulic pressure is supplied by a pump through the inlet seal and air is purged through the outlet seal which then closes. The higher the pressure, the greater the sealing strength of the seal which prevents accidental unplugging. There are different types of seals: for interior plain-ended openings, for pipes with plain-ended opening, for flanged pipes. (author)

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

Energy Technology Data Exchange (ETDEWEB)

Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

2012-08-15

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 Degree-Sign C, cooling to 522 Degree-Sign C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at {approx}0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher {Delta}T between fuel center and cladding than at core center, together providing more rare earths at the cladding and

Fuel assembly duct cutting in the FFTF/IEM Cell

International Nuclear Information System (INIS)

Gibbons, P.W.

1985-01-01

Two mill type slitting cutters are used in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell during the disassembly sequence of a Driver Fuel Assembly. This disassembly is necessary so that selected parts may be examined both in the IEM Cell and elsewhere. The cutters have been in use for two years. During this time eight Driver Fuel assemblies have been taken apart in the IEM Cell. The cutters' operating philosophy and characteristics, as well as lessons learned from a significant equipment failure are presented. 1 ref., 6 figs., 1 tab

Accommodation of potential hydrogen formation in LMFBR accidents

International Nuclear Information System (INIS)

Stepnewski, D.D.; Peak, R.D.; Mahaffey, M.K.

1981-01-01

Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus avoiding the formation of hydrogen. When liner failures are postulate in hypothetical reactor vessel meltthrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed

Standard Test Method for Testing Polymeric Seal Materials for Geothermal and/or High Temperature Service Under Sealing Stress

CERN Document Server

American Society for Testing and Materials. Philadelphia

1985-01-01

1.1 This test method covers the initial evaluation of (screening) polymeric materials for seals under static sealing stress and at elevated temperatures. 1.2 This test method applies to geothermal service only if used in conjunction with Test Method E 1068. 1.3 The test fluid is distilled water. 1.4 The values stated in SI units are to be regarded as the standard. The values in parentheses are for information only. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Microstructural examination of 12% Cr martensitic stainless steel after irradiation at elevated temperatures in FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Hsu, Chen-Yih; Gelles, D.S.; Lechtenberg, T.A.

1986-06-01

A remelted 12% Cr martensitic stainless steel (HT-9) has been examined by transmission electron microscopy before and after irradiation in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The irradiation temperatures were 365,420, 520, and 600 degree C with the fluences as high as 7.3 x 10 22 n/cm 2 (E > 0.1 MeV) or 34 dpa. The extracted precipitates from each specimen were identified using x-ray microanalysis and selected area diffraction. The precipitates in the unirradiated condition were primarily M 23 C 6 carbides, which formed at martensite lath and prior austenite grain boundaries. During irradiation at elevated temperatures, small amounts of other phases formed, which were tentatively identified as the chromium-rich α', the nickel-silicon rich G-phase, and the intermetallic Chi phase. Irradiation-induced voids were observed only in specimens irradiated at 420 degree C to a dose of 34 dpa; no voids were found for specimens irradiated at 365, 520, and 600 degree C (∼11, ∼34, and ∼34 dpa). These results are not in agreement with previous experiments in that voids have not been reported in this alloy at relatively high fluence level (∼67 dpa) following irradiation in another fast-spectrum reactor (EBR.II). This is, however, the first observation following FFTF irradiation. The present results indicate that cavities can form in HT-9 at modest fluence levels even without significant generation of helium. Hence, the cavity formation in this class of ferritic alloys is not simply caused by helium generation but rather more complex mechanisms. 12 refs., 2 figs., 3 tabs

Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

International Nuclear Information System (INIS)

Hilliard, R.K.; Postma, A.K.

1975-01-01

Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

CEC activities in the field of LMFBR safety

International Nuclear Information System (INIS)

Balz, W.; Finzi, S.; Klersy, R.

1976-01-01

The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

Thermo-Mechanical Test of Seal System in Flexible Pipe End Fittings

DEFF Research Database (Denmark)

Banke, Lars

1999-01-01

are driven radially into the barrier layer and supported by the surrounding steel casing. In order to verify the integrity of the concept the seal system is subjected cyclic pressure and temperature variations to simulate the service conditions.The aim of the testing is to demonstrate the sensitivity...... of the seal system geometry and its tolerances necessary to maintain a tight seal. The test is carried out in a purpose built autoclave, in which the seal system can be tested while undergoing variations in pressure and temperature.The paper will present a study on the importance of the geometry of the gasket...... and the inner liner. The inner and outer diameter of the gasket are varied to see the effectiveness of the seal mechanism. The effect of varying the width of the gasket as well as the surface roughness of the components in the seal system is analysed. Finally, it is investigated how the seal system is affected...

FFTF vertical sodium storage tank preliminary thermal analysis

International Nuclear Information System (INIS)

Irwin, J.J.

1995-01-01

In the FFTF Shutdown Program, sodium from the primary and secondary heat transport loops, Interim Decay Storage (IDS), and Fuel Storage Facility (FSF) will be transferred to four large storage tanks for temporary storage. Three of the storage tanks will be cylindrical vertical tanks having a diameter of 28 feet, height of 22 feet and fabricated from carbon steel. The fourth tank is a horizontal cylindrical tank but is not the subject of this report. The storage tanks will be located near the FFTF in the 400 Area and rest on a steel-lined concrete slab in an enclosed building. The purpose of this work is to document the thermal analyses that were performed to ensure that the vertical FFTF sodium storage tank design is feasible from a thermal standpoint. The key criterion for this analysis is the time to heat up the storage tank containing frozen sodium at ambient temperature to 400 F. Normal operating conditions include an ambient temperature range of 32 F to 120 F. A key parameter in the evaluation of the sodium storage tank is the type of insulation. The baseline case assumed six inches of calcium silicate insulation. An alternate case assumed refractory fiber (Cerablanket) insulation also with a thickness of six inches. Both cases assumed a total electrical trace heat load of 60 kW, with 24 kW evenly distributed on the bottom head and 36 kW evenly distributed on the tank side wall

FFTF-cycle 10 program and future plan

Science.gov (United States)

Kohyama, Akira

1988-04-01

Brief outlines are provided of the FFTF cycle 10 program and future plans in consideration. The primary objective of the Japan-US collaboration program is to enable predictions of material behavior in MFRs to be made from data obtained in other irradiation environments. Major program goals are outlined.

SASSYS LMFBR systems code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

1983-01-01

The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

Electronically controlled mechanical seal for aerospace applications--Part 2: Transient tests

Science.gov (United States)

Wolff, Paul J.; Salant, Richard F.

1995-01-01

An electronically controlled mechanical seal for use as the purge gas seal in a liquid oxygen turbopump has been fabricated and tested under transient operating conditions. The thickness of the lubricating film is controlled by adjusting the coning of the carbon face. This is accomplished by applying a voltage to a piezoelectric actuator to which the carbon face is bonded. The seal has been operated with a closed-loop control system that utilizes either the leakage rate or seal face temperature as the feedback. Both speed and pressure transients have been imposed on the seal. The transient tests have demonstrated that the seal is capable of maintaing low leakage rates while limiting face temperatures.

Research and development of bellows for LMFBR in Japan

International Nuclear Information System (INIS)

Takahashi, T.; Mukai, K.; Yamamoto, K.

1980-01-01

Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

Research and development of bellows for LMFBR in Japan

International Nuclear Information System (INIS)

Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

1979-11-01

The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

Knowledge Management at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Wootan, David; Omberg, Ron

2013-01-01

Conclusions: • Accessibility of information from FFTF has been substantially increased; • FFTF Startup Test information available – FFTF had a disciplined, organized approach; • Extensive Startup Characterization program data available; • Extensive Passive safety testing data available

An assessment of the viability of storing FFTF sodium in tank cars

International Nuclear Information System (INIS)

Young, M.W.; Burke, T.M.

1995-01-01

Current FFTF Transition Project plans call for construction of a Sodium Storage Facility to store the plant sodium until it is processed either as product or waste. This report evaluates an alternative concept which would store the sodium in rail tank cars. It is concluded that utilizing a simple facility for offloading the FFTF sodium to standard industrial tank cars is not technically viable. Mitigation of potential radioactive sodium spills requires that the offload facility incorporate many of the features of the sodium storage facility. With these mitigation features incorporated, there is no significant cost or schedule advantage for the option of storing the FFTF sodium in tank cars when compared to the currently planned SSF. In addition, it is believed that the tank car option results in higher risk to project success because of unknowns associated with technical, regulatory, and public perception issues. It is therefore recommended that the project proceed with definitive design of the SSF

Small leak shutdown, location, and behavior in LMFBR steam generators

International Nuclear Information System (INIS)

Sandusky, D.W.

1976-01-01

The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

International Nuclear Information System (INIS)

Richard, R.F.

1995-01-01

It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

International Nuclear Information System (INIS)

Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

1985-04-01

The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

Practical considerations from construction of the FFTF pipe support system

International Nuclear Information System (INIS)

Hartsell, J.O.

1979-04-01

Late this year the FFTF is scheduled to initiate criticality and power operation, an event signifying complete resolution of construction problems and compliance with applicable requirements. One of the tasks accomplished to achieve this position was the conduct of cell completion walkthroughs, problem identification, and resolution. Heavily oriented toward the final verification of pipe supports, the cell completion process identified and resolved a series of problems not readily apparent during the detailed design phases of plant construction. It is the objective of this presentation to establish a degree of awareness of these types of problems and of the FFTF techniques employed in resolution. All of the problems depicted herein have been solved and the installations corrected

LMFBR accident delineation study: approach and preliminary results

International Nuclear Information System (INIS)

Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

1979-01-01

Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

Quality Assurance Program Plan for FFTF effluent controls. Revision 1

International Nuclear Information System (INIS)

Seamans, J.A.

1995-01-01

This Quality Assurance Program Plan is specific to environmental related activities within the FFTF Property Protected Area. The activities include effluent monitoring and Low Level Waste Certification

Sealing performance test for main flange of pressure vessel of T2 test section in HENDEL

International Nuclear Information System (INIS)

Ioka, Ikuo; Inagaki, Yoshiyuki; Matsumoto, Kiminori; Kondou, Yasuo; Suzuki, Kunihiko; Miyamoto, Yoshiaki; Asami, Masanobu.

1990-12-01

A pressure vessel of T 2 test section in helium engineering demonstration loop (HENDEL) was fabricated to the same scale of the reactor pressure vessel made of 2(1/4)Cr-1Mo steel in high temperature engineering test reactor (HTTR). Also, the sealing structure of a main flange of pressure vessel in T 2 test section was composed of the double metal O-rings and Ω-seal which would be used in the sealing structure of HTTR. The sealing performance test for the main flange of the pressure vessel in T 2 test section was carried out to confirm the integrity of sealing structure of a main flange in HTTR. T 2 test section has been operated about 7700 hours in previous 18 cycles. The leakage of helium gas from inner metal O-ring was measured by the static pressurized process under the operating condition of HTTR (helium gas: 400degC, 40kg/cm 2 G, 4gk/s). The calculated leakage of helium gas was less than 9.6x10 -7 atm·cm 3 /sec. From the result, it is expected that the sealing structure of main flange in HTTR would maintain the leak tightness in the life. (author)

Tamper-indicating devices and safeguards seals evaluation test report. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Horton, P.R.V.; Waddoups, I.G.

1995-08-01

Volume I was based on a survey and an evaluation of seals that are used as tamper-indicating devices at DOE facilities. For that evaluation, currently available seals were physically and environmentally evaluated under two broad categories: handling durability and tamper resistance. Our study indicated that the environmental testing had no negative effects on the results of the mechanical tests. In Volume II, we evaluate some loop, fiber optic loop, and pressure-sensitive seals that are not used at DOE facilities. However, we continue to focus on qualities required by DOE: durability and tamper resistance. The seals are comparatively rated, and recommendations are made for using currently available seals and new tamper-indicating device technology.

Tamper-indicating devices and safeguards seals evaluation test report. Volume 2

International Nuclear Information System (INIS)

Horton, P.R.V.; Waddoups, I.G.

1995-08-01

Volume I was based on a survey and an evaluation of seals that are used as tamper-indicating devices at DOE facilities. For that evaluation, currently available seals were physically and environmentally evaluated under two broad categories: handling durability and tamper resistance. Our study indicated that the environmental testing had no negative effects on the results of the mechanical tests. In Volume II, we evaluate some loop, fiber optic loop, and pressure-sensitive seals that are not used at DOE facilities. However, we continue to focus on qualities required by DOE: durability and tamper resistance. The seals are comparatively rated, and recommendations are made for using currently available seals and new tamper-indicating device technology

Analytical approach for confirming the achievement of LMFBR reliability goals

International Nuclear Information System (INIS)

Ingram, G.E.; Elerath, J.G.; Wood, A.P.

1981-01-01

The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability

Testing of ethylene propylene seals for the GA-4/GA-9 casks

International Nuclear Information System (INIS)

Boonstra, R.H.

1993-01-01

The primary O-ring seal of the GA-4 and GA-9 casks was tested for leakage with a full-scale mockup of the cask lid and flange. Tests were performed at temperatures of ambient, -41, 121, and 193 C. Shim plates between the lid and flange simulated gaps caused by thermal distortion. The testing used a helium mass spectrometer leak detector (MSLD). Results showed that the primary seal was leaktight for all test conditions. Helium permeation through the seal began in 13--23 minutes for the ambient tests and in 1--2 minutes for the tests at elevated temperatures. After each test several hours of pumping were typically required to reduce the MSLD background reading to an acceptable level for the next test, indicating that the seal had become saturated with helium. To verify that the test results showed permeation and not real leakage, several response checks were conducted in which a calibrated leak source was inserted in the detector line near the seal. When the leak source was activated the detector responded within seconds

Testing of ethylene propylene seals for the GA-4/GA-9 casks

International Nuclear Information System (INIS)

Boonstra, R.H.

1993-01-01

The primary O-ring seal of the GA-4 and GA-9 casks was tested for leakage with a full-scale mockup of the cask lid and flange. Tests were performed at temperatures of ambient, -41 degrees, 121 degrees, and 193 degrees C. Shim plates between the lid and flange simulated gaps caused by thermal distortion. The testing used a helium mass spectrometer leak detector (MSLD). Results showed that the primary seal was leaktight for all test conditions. Helium permeation through the seal began in 13--23 minutes for the ambient tests and in 1--2 minutes for the tests at elevated temperatures. After each test several hours of the pumping were typically required to reduce the MSLD background reading to an acceptable level for the next test, indicating that the seal had become saturated with helium. To verify that the test results showed permeation and not real leakage, several response checks were conducted in which a calibrated leak source was inserted in the detector line near the seal. When the leak source was activated the detector responded within seconds

FFTF-containment air-cleaning system

International Nuclear Information System (INIS)

Mahaffey, M.K.; Stepnewski, D.D.

1981-01-01

The FFTF Containment can accommodate all design basis events and the hypothetical core disruptive accident with adequate margin without a venting or purging system; however, in concert with the development objective, a system was designed and constructed to evaluate technology related to containment atmosphere venting and cleanup functions. The system can be used to purge high H 2 concentrations or to vent excessive containment pressure. In either case containment atmosphere is exhausted through an aqueous scrubber system consisting of a venturi scrubber and fibrous filter bank

LMFBR plant parameters

International Nuclear Information System (INIS)

1985-07-01

This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

Technical considerations relative to removal of sodium from LMFBR components

Energy Technology Data Exchange (ETDEWEB)

McDonald, J S; Asquith, J G

1975-07-01

Reviewed in this paper are technical considerations which are of importance in choosing between an alcohol process and a moist nitrogen process for the removal of sodium from LMFBR components. Results observed in laboratory tests and in the cleaning of large scale components (e.g. a 28 MWt Modular Steam Generator Test Unit) are presented and discussed. (author)

Comments on US LMFBR steam generator base technology

International Nuclear Information System (INIS)

Simmons, W.R.

1984-01-01

The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

Instructions for submittal and control of FFTF design documents and design related documentation

International Nuclear Information System (INIS)

Grush, R.E.

1976-10-01

This document provides the system and requirements for management of FFTF technical data prepared by Westinghouse Hanford (HEDL), and design contractors, the construction contractor and lower tier equipment suppliers. Included in this document are provisions for the review, approval, release, change control, and accounting of FFTF design disclosure and base documentation. Also included are provisions for submittal of other design related documents for review and approval consistent with applicable requirements of RDT-Standard F 2-2, ''Quality Assurance Program Requirements.''

Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

International Nuclear Information System (INIS)

Shin, Y.S.; Wambsganss, M.W.

1975-05-01

This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Allard, L.T.; Elerath, J.G.

1976-02-01

This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

LMFBR flexible pipe joint development program. Annual technical progress report, government fiscal year 1977

International Nuclear Information System (INIS)

1978-01-01

Currently, the ASME Boiler and Pressure Vessel Code does not allow the use of flexible pipe joints (bellows) in Section III, Class 1 reactor primary piping systems. Studies have shown that the primary piping loops of LMFBR's could be simplified by using these joints. This simplification translates directly into shorter primary piping runs and reduced costs for the primary piping system. Further cost savings result through reduced vault sizes and reduced containment building diameter. In addition, the use of flexible joints localizes the motions from thermally-induced piping growth into components which are specifically designed to accommodate this motion. This reduces the stress levels in the piping system and its components. It is thus economically and structurally important that flexible piping joints be available to the LMFBR designer. The overall objective of the Flexible Joint Program is to provide this availability. This will be accomplished through the development of ASME rules which allow the appropriate use of such joints in Section III, Class 1 piping systems and through the development and demonstration of construction methods which satisfy these rules. The rule development includes analytic and testing methodology formulations which will be supported by subscale bellows testing. The construction development and demonstration encompass the design, fabrication, and in-sodium testing of prototypical LMFBR plant-size flexible pipe joints which meet all ASME rule requirements. The satisfactory completion of these developmental goals will result in an approved flexible pipe joint design for the LMFBR. Progress is summarized in the following efforts undertaken during 1977 to accomplish these goals: (1) code case support, (2) engineering and design, (3) material development, (4) testing, and (5) manufacturing development

Assessment of helium effects on swelling by reirradiation in FFTF of Path A alloys previously irradiated in HFIR

International Nuclear Information System (INIS)

Maziasz, P.J.; Garner, F.A.; Brager, H.R.

1985-01-01

Specimens of the Path A Prime Candidate Alloys and of N-lot SS 316 were irradiated in HFIR at 400 to 600 0 C to fluences producing approximately 10 to 44 dpa and 500 to 3600 at. ppm He, in both the solution annealed and 20 to 25% cold-worked conditions. The cavity swelling and total microstructural evolution of most samples were observed via transmission electron microscopy on identical disks irradiated side by side in HFIR, and immersion densities were also measured prior to insertion into FFTF/MOTA (Materials Open Test Assembly of the Fast Flux Test Facility). These disks are being irradiated in the FFTF/MOTA (cycles 5 and 6), side by side with disks of the same materials which were not previously irradiated in HFIR. These specimens have been divided into two subsets for discharges after 30 and 60 dpa. 4 references, 1 table

10 CFR 34.27 - Leak testing and replacement of sealed sources.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Leak testing and replacement of sealed sources. 34.27... SAFETY REQUIREMENTS FOR INDUSTRIAL RADIOGRAPHIC OPERATIONS Equipment § 34.27 Leak testing and replacement... radiographic exposure device and leak testing of any sealed source must be performed by persons authorized to...

Clay sealing for dumps - new findings from suitability tests

Energy Technology Data Exchange (ETDEWEB)

Kunde, L.; Kuntsche, K.; Duellmann, H.

1988-06-01

Extensive suitability tests were carried out in the laboratory and field on various clays of the Lower-Rhineland brown coal mining area. The purpose of these tests was to determine the basic suitability of these clays for use as sealing layers for residue dumps. The tests also furnished criteria according to which a specific imperviousness can be achieved. The results of the tests can be summed up as follows: Since the material properties of the in-situ clays vary widely, not only mean values but also corresponding variation coefficients must be adhered to. For the purpose of checking the quality - contrary to the hitherto applied assessment practice - the decisive factor is not the testing of the dry density as compared to the Proctor density, but the homogeneity of the sealing layer that is achieved as a result of the size reduction and compaction of the pseudo-granular structure. Using suitable equipment, homogeneities, for which permeability coefficients of k less than or equal to 10/sup -9/ m/s in the dump sealing are obtained can be achieved with the clays that were examined.

FFTF report: FFTF piping installation and welding techniques

International Nuclear Information System (INIS)

Gilles, J.

1975-01-01

The main sodium piping with a diameter of 16'' or 28 '' is being installed at the FFTF construction site starting in December 1974. The supplier and authority demarcations are: Combustion Engineering supplies the reactor vessel, guard vessel and adjoining pipes and uses the machine welding equipment ''Dimetrics''; for the piping system of the primary and secondary loops the pipes manufactured by Rollmet at HUICO, Pasco, were delivered and prefabricated there, as far as compatible with the installation. ''Astroarc'' welding machines are used by Bechtel for the piping prefabrication in the weld laboratory as well as on site at the construction site. Technical welding problems occurring during the course of the installation at the construction site and several during this time are described. At present 6 weld seams in the reactor and 14 weld seams in the secondary loop are accepted. The requirement exists to carry out as many welds as possible automatically, in order to produce sodium pipe welds of high technical quality and which are reproducible. The welding equipment is described

Proposed sealing field tests for a potential high-level radioactive waste repository in unsaturated tuff

International Nuclear Information System (INIS)

Fernandez, J.A.; Case, J.B.; Tyburski, J.

1992-01-01

This paper contains a general description of the field tests proposed for the Yucca Mountain Site Characterization Project repository sealing program. The tests are intended to resolve various performance and emplacement concerns associated with sealing components. Ten discrete tests are proposed to address these concerns. These tests are divided into two categories -- simple and complex tests. The simple tests are: the small-scale in situ tests: the intermediate-scale borehole seal tests; the fracture grouting tests; the surface backfill tests; and the grouted rock mass tests. The complex tests are the seepage control tests; in situ backfill tests; in situ bulkhead tests; large-scale shaft seal tests; and remote borehole seal tests. These tests are proposed to be performed in welded and nonwelded tuff environments. The final selection of sealing tests will depend on the nature of the geologic and hydrologic conditions encountered during the development of the exploratory studies facility. Some tests may be performed before license application and some after license application

Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

International Nuclear Information System (INIS)

Broc, Daniel

2001-01-01

As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

Analysis of carbon transport in the EBR-II and FFTF primary sodium systems

International Nuclear Information System (INIS)

Snyder, R.B.; Natesan, K.; Kassner, T.F.

1976-01-01

An analysis of the carburization-decarburization behavior of austenitic stainless steels in the primary heat-transport systems of the EBR-II and FFTF has been made that is based upon a kinetic model for the diffusion process and the surface area of steel in contact with flowing sodium at various temperatures in the two systems. The analysis was performed for operating conditions that result in sodium outlet temperatures of 474 and 566 0 C in the FFTF and 470 0 C in the EBR-II. If there was no external source of carbon to the system, i.e., other than the carbon initially present in the steel and the sodium, the dynamic-equilibrium carbon concentrations calculated for the FFTF primary sodium were approximately 0.025 and approximately 0.065 ppm for the 474 and 566 0 C outlet temperatures, respectively, and approximately 0.018 ppm for the EBR-II primary system. The analysis indicated that a carbon-source rate of approximately 250 g/y would be required to increase the carbon concentration of the EBR-II sodium to the measured range of approximately 0.16--0.19 ppm. An evaluation of possible carbon sources and the amount of carbonaceous material introduced into the reactor cover gas and sodium suggests that the magnitude of the calculated contamination rate is reasonable. For a 566 0 C outlet temperature, carbonaceous material would have to be introduced into the FFTF primary system at a rate approximately 4--6 times higher than in EBR-II to achieve the same carbon concentration in the sodium in the two systems. Since contamination rates of approximately 1500 g/y are unlikely, high-temperature fuel cladding in the FFTF should exhibit decarburization similar to that observed in laboratory loop systems, in contrast to the minimal compositional changes that result after exposure of Type 316 stainless steel to EBR-II sodium at temperatures between approximately 625 and 650 0 C

Aluminized film, seam sealing tests and observations. Final report

Energy Technology Data Exchange (ETDEWEB)

1994-06-16

The purpose of this work was to investigate various seam sealing techniques, reinforcing methods, fitting installations, seam tolerances and geometric configurations pertinent to an aluminized plastic laminate. The program seeks a successful fabricating method for producing low-diffusion, cylindrical, spar liners to contain pressurized GH{sub 2} and GO{sub 2}. The test plan included: (1) seaming techniques on metallized Mylar film; (2) ``double patches`` for end fittings; (3) stainless steel bulkhead fitting assembly with seals; (4) minimum run tolerance on linear shear seam; (5) peel seam vs. inverted seal seam fabrication.

Man-machine enhancements to existing FFTF control panels

International Nuclear Information System (INIS)

Miller, E.M.

1984-01-01

FFTF Project enhanced existing control panels with tape and labels to mitigate operator problems and to incorporate the guidance of NUREG-0700. The enhancements grouped displays and controls into meaningful units and labelled controls and displays to facilitate their identification and efficient use. The enhancements were inexpensive and well received by the facility's operations staff

Structural and containment response to LMFBR accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

1978-01-01

The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

LMFBR thermal-striping evaluation

International Nuclear Information System (INIS)

Brunings, J.E.

1982-10-01

Thermal striping is defined as the fluctuating temperature field that is imposed on a structure when fluid streams at different temperatures mix in the vicinity of the structure surface. Because of the uncertainty in structural damage in LMFBR structures subject to thermal striping, EPRI has funded an effort for the Rockwell International Energy Systems Group to evaluate this problem. This interim report presents the following information: (1) a Thermal Striping Program Plan which identifies areas of analytic and experimental needs and presents a program of specific tasks to define damage experienced by ordinary materials of construction and to evaluate conservatism in the existing approach; (2) a description of the Thermal Striping Test Facility and its operation; and (3) results from the preliminary phase of testing to characterize the fluid environment to be applied in subsequent thermal striping damage experiments

Japanese materials program and FFTF

International Nuclear Information System (INIS)

Ishino, Shiori

1988-01-01

Japanese materials program has been briefly reviewed and the associated university program, which is still in a provisional stage has been described in some detail. Important elements of the university proposal will be 1) construction of a high energy high fluence neutron irradiation facility, 2) establishing or expanding local research centers including hot laboratories, and 3) promotion of fundamental studies. The FFTF/MOTA Project is a very important constituent of the whole program, the results coming out of which should be well coordinated with other fundamental research programs to extract full essence needed for the advancement of realization of fusion energy. (author)

Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

1997-04-01

Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2015-01-01

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

Design of Fire/Gas Penetration Seals and fire exposure tests for Tokamak Fusion Test Reactor experimental areas

International Nuclear Information System (INIS)

Cavalluzzo, S.

1983-01-01

A Fire/Gas Penetration Seal is required in every penetration through the walls and ceilings into the Test Cell housing the Tokamak Fusion Test Reactor (TFTR), as well as other adjacent areas to protect the TFTR from fire damage. The penetrations are used for field coil lead stems, diagnostics systems, utilities, cables, trays, mechanical devices, electrical conduits, vacuum liner, air conditioning ducts, water pipes, and gas pipes. The function of the Fire/Gas Penetration Seals is to prevent the passage of fire and products of combustion through penetrations for a period of time up to three hours and remain structurally intact during fire exposure. The Penetration Seal must withstand, without rupture, a fire hose water stream directed at the hot surface. There are over 3000 penetrations ranging in size from several square inches to 100 square feet, and classified into 90 different types. The material used to construct the Fire/Gas Penetration Seals consist of a single and a two-component room temperature vulcanizing (RTV) silicone rubber compound. Miscellaneous materials such as alumina silica refractory fibers in board, blanket and fiber forms are also used in the construction and assembly of the Seals. This paper describes some of the penetration seals and the test procedures used to perform the three-hour fire exposure tests to demonstrate the adequacy of the seals

Future development LMFBR-steam generators SNR2

International Nuclear Information System (INIS)

Essebaggers, J.; Pors, J.G.

1975-01-01

The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

International Nuclear Information System (INIS)

Bordner, G.L.

1979-10-01

The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

Leakage and Power Loss Test Results for Competing Turbine Engine Seals

National Research Council Canada - National Science Library

Proctor, Margaret

2004-01-01

.... To address engine manufacturers' concerns about the heat generation and power loss from these contacting seals, brush, finger, and labyrinth seals were tested in the NASA High Speed, High Temperature...

{open_quotes}O{close_quotes} ring sealed process tube, Phase II, test project

Energy Technology Data Exchange (ETDEWEB)

Johnson, R.E.

1951-04-09

The {open_quotes}O{close_quotes} ring seal has been proposed to replace the van stone flange and the bellows thermal expansion assembly currently used on the existing Hanford piles to achieve water and gas seals, respectively. Possible advantages of the {open_quotes}O{close_quotes} ring seal are: (1) simplification of component parts and elimination of van stone corrosion; (2) simplification of maintenance; (3) lower costs of initial erection; (4) increased strength. This test supplements Test Project No. 27 (a preliminary thermal cycling test) in applying the {open_quotes}O{close_quotes} ring seal assembly to actual pile operating conditions.

Prototype bellows sealed nuclear valve development -reliability through testing

International Nuclear Information System (INIS)

Dixon, D.F.; Abbas, M.

1978-01-01

To assist in appraising bellows sealed valve performance, 10 tests were done on a ''1 in.'' prototype bellows sealed valve design. The tests simulated primary heat transport (PHT) system conditions for a 600 MWe CANDU-PHW. The design approach was to have all valve components outlast the bellows in endurance tests; this was achieved. The valve design meets the Atomic Energy of Canada Limited specification. For comparison, bellows fatigue failure data were fitted to both log-normal and Weibull distributions. A numerical example shows how to select valve stroke amplitude on the basis of valve flow requirement and the minimum acceptable fatigue life. (author)

Design and safety evaluation of radioactive gas handling and storage in the FFTF

International Nuclear Information System (INIS)

Armstrong, G.R.; Hale, J.P.; Halverson, T.G.

1976-01-01

During the operation of the Fast Flux Test Facility (FFTF), radioactive gases, primarily xenon and krypton, will be produced which will require processing and storing. Two systems have been installed in the FFTF for handling these gases: (1) one to handle, primarily, the reactor cover gas system, and (2) a second to handle the cells and cover gas systems, other than the reactor, whose atmosphere may become contaminated. The system that processes the reactor cover gas, which is argon, is called the Radioactive Argon Processing System (RAPS). The effluent argon from RAPS will normally be sufficiently decontaminated to allow its reuse as the reactor cover gas. If the radioactive level in the RAPS becomes too high, the exhaust stream will be diverted to the Cell Atmosphere Processing System (CAPS), a system which can function as a backup to RAPS. The design and operation of the RAPS and CAPS systems are described and certain safety aspects of the systems are discussed. It is shown that these systems adequately provide the cleanup services required and that they provide the safety margins necessary to assure adequate safety to the public

Fuels and materials testing capabilities in Fast Flux Test Facility

International Nuclear Information System (INIS)

Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

1989-01-01

The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

Hydrogen formation and control under postulated LMFBR accident conditions

International Nuclear Information System (INIS)

Armstrong, G.R.; Wierman, R.W.

1976-09-01

The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

Mode of failure of LMFBR fuel pins

International Nuclear Information System (INIS)

Washburn, D.F.

1975-01-01

The objectives of the irradiation test described were to evaluate mixed-oxide fuel performance and to confirm the design adequacy of the FFTF fuel pins. After attainment of the initial objectives the irradiation of several of the original fuel pins was continued until a cladding breach occurred. The consequences of a cladding breach were evaluated by reconstituting the original 37-pin subassembly into two 19-pin subassemblies after a burnup at 50,000 MWd/MTM (5.2 a/o). The original pins were supplemented with fresh pins as necessary. Irradiation of the subassemblies was continued until a cladding breach occurred. Results are presented and discussed

Field Test and Evaluation of Engineered Biomineralization Technology for Sealing Existing Wells

Energy Technology Data Exchange (ETDEWEB)

Cunningham, Alfred [Montana State Univ., Bozeman, MT (United States)

2015-12-21

This research project addresses one of the goals of the U.S. Department of Energy (DOE) Carbon Storage Program (CSP) aimed at developing Advanced Wellbore Integrity Technologies to Ensure Permanent Geologic Carbon Storage. The technology field-tested in this research project is referred to as microbially induced calcite precipitation (MICP), which utilizes a biologically-based process to precipitate calcium carbonate. If properly controlled MICP can successfully seal fractures, high permeability zones, and compromised wellbore cement in the vicinity of wellbores and in nearby caprock, thereby improving the storage security of geologically-stored carbon dioxide. This report describes an MICP sealing field test performed on a 24.4 cm (9.625 inch) diameter well located on the Gorgas Steam Generation facility near Jasper, Alabama. The research was aimed at (1) developing methods for delivering MICP promoting fluids downhole using conventional oil field technologies and (2) assessing the ability of MICP to seal cement and formation fractures in the near wellbore region in a sandstone formation. Both objectives were accomplished successfully during a field test performed during the period April 1-11, 2014. The test resulted in complete biomineralization sealing of a horizontal fracture located 340.7 m (1118 feet) below ground surface. A total of 24 calcium injections and six microbial inoculation injections were required over a three day period in order to achieve complete sealing. The fractured region was considered completely sealed when it was no longer possible to inject fluids into the formation without exceeding the initial formation fracture pressure. The test was accomplished using conventional oil field technology including an 11.4 L (3.0 gallon) wireline dump bailer for injecting the biomineralization materials downhole. Metrics indicating successful MICP sealing included reduced injectivity during seal formation, reduction in pressure falloff, and

Intermediate-Size Inducer Pump design report. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Boardman, T.J.

1979-06-15

This report summarizes the mechanical, structural, and hydrodynamic design of the Intermediate-Size Inducer Pump (ISIP). The design was performed under Atomics International's DOE Base Technology Program by the Atomics International and Rocketdyne Divisions of Rockwell International. The pump was designed to utilize the FFTF prototype pump frame as a test vehicle to test the inducer, impeller, and diffuser plus necessary adapter hardware under simulated Large Scale Liquid Metal Fast Breeder Reactor service conditions. The report describes the design requirements including the purpose and objectives, and discusses those design efforts and considerations made to meet the requirements. Included in the report are appendices showing calculative methods and results. Also included are overall assembly and layout drawings plus some details used as illustrations for discussion of the design results and the results of water tests performed on a model of the inducer.

FFTF primary heat transport system heating, ventilating and air conditioning system experience

International Nuclear Information System (INIS)

Umek, A.M.; Hicks, D.F.; Schweiger, D.L.

1981-01-01

FFTF cools its primary/in-containment sodium equipment cells by means of a forced nitrogen cooling system which exchanges heat with a water-glycol system. The nitrogen cooling system is also used to maintain an inert gas atmosphere in the cells containing sodium equipment. Sodium Piping and Components have installed electrical resistance heaters to maintain a minimum sodium temperature and stainless steel jacketed mineral insulation to reduce heat loss. Design features and test results of a comprehensive redesign of the HVAC and insulation system required to support long-term nuclear operations are discussed

Studies of LMFBR: method of analysis and some results

International Nuclear Information System (INIS)

Ishiguro, Y.; Dias, A.F.; Nascimento, J.A. do.

1983-01-01

Some results of recent studies of LMFBR characteristics are summarized. A two-dimensional model of the LMFBR is taken from a publication and used as the base model for the analysis. Axial structures are added to the base model and a three-dimensional (Δ - Z) calculation has been done. Two dimensional (Δ and RZ) calculations are compared with the three-dimensional and published results. The eigenvalue, flux and power distributions, breeding characteristics, control rod worth, sodium-void and Doppler reactivities are analysed. Calculations are done by CITATION using six-group cross sections collapsed regionwise by EXPANDA in one-dimensional geometries from the 70-group JFS library. Burnup calculations of a simplified thorium-cycle LMFBR have also been done in the RZ geometry. Principal results of the studies are: (1) the JFS library appears adequate for predicting overall characteristics of an LMFBR, (2) the sodium void reactivity is negative within - 25 cm from the outer boundary of the core, (3) the halflife of Pa-233 must be considered explicitly in burnup analyses, and (4) two-dimensional (RZ and Δ) calculations can be used iteratively to analyze three-dimensional reactor systems. (Author) [pt

Acoustics and voiding dynamics during SLSF simulations of LMFBR undercooling transients

International Nuclear Information System (INIS)

Anderson, T.T.; Kuzay, T.M.; Marr, W.W.; Miles, K.J.; Pedersen, D.R.; Thompson, D.H.; Wilson, R.E.

1978-01-01

The SLSF is the largest U.S. in-reactor test vehicle for steady-state and transient experiments in an environment typical of a LMFBR core. The SLSF experiment program, sponsored by the Department of Energy, contributes to the LMFBR safety assurance program by providing data on key phenomena that occur during postulated reactor accidents. This paper describes completed SLSF experiments, in-core instrumentation used, and methods of data interpretation to determine sodium boiling and voiding dynamics. Boiling inception is shown to be identifiable from several types of in-core instruments. Location of the boiling front and void growth derived from experimental data are compared with analytical predictions. These and other data form the basis to improve understanding of accidents and to validate or guide the development of accident analysis methods

Ceramic Seal.

Energy Technology Data Exchange (ETDEWEB)

Smartt, Heidi A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Romero, Juan A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Custer, Joyce Olsen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hymel, Ross W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Krementz, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gobin, Derek [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Harpring, Larry [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martinez-Rodriguez, Michael [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Varble, Don [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DiMaio, Jeff [Tetramer Technologies, Pendleton, SC (United States); Hudson, Stephen [Tetramer Technologies, Pendleton, SC (United States)

2016-11-01

Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.

Ceramic Seal

International Nuclear Information System (INIS)

Smartt, Heidi A.; Romero, Juan A.; Custer, Joyce Olsen; Hymel, Ross W.; Krementz, Dan; Gobin, Derek; Harpring, Larry; Martinez-Rodriguez, Michael; Varble, Don; DiMaio, Jeff; Hudson, Stephen

2016-01-01

Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.

A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

International Nuclear Information System (INIS)

Hartung, J.; Berk, S.

1978-01-01

Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

Vacuum seals design and testing for a linear accelerator of the National Spallation Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Z. Chen; C. Gautier; F. Hemez; N. K. Bultman

2000-02-01

Vacuum seals are very important to ensure that the Spallation Neutron Source (SNS) Linac has an optimum vacuum system. The vacuum joints between flanges must have reliable seals to minimize the leak rate and meet vacuum and electrical requirements. In addition, it is desirable to simplify the installation and thereby also simplify the maintenance required. This report summarizes an investigation of the metal vacuum seals that include the metal C-seal, Energized Spring seal, Helcoflex Copper Delta seal, Aluminum Delta seal, delta seal with limiting ring, and the prototype of the copper diamond seals. The report also contains the material certifications, design, finite element analysis, and testing for all of these seals. It is a valuable reference for any vacuum system design. To evaluate the suitability of several types of metal seals for use in the SNS Linac and to determine the torque applied on the bolts, a series of vacuum leak rate tests on the metal seals have been completed at Los Alamos Laboratory. A copper plated flange, using the same type of delta seal that was used for testing with the stainless steel flange, has also been studied and tested. A vacuum seal is desired that requires significantly less loading than a standard ConFlat flange with a copper gasket for the coupling cavity assembly. To save the intersegment space the authors use thinner flanges in the design. The leak rate of the thin ConFlat flange with a copper gasket is a baseline for the vacuum test on all seals and thin flanges. A finite element analysis of a long coupling cavity flange with a copper delta seal has been performed in order to confirm the design of the long coupling cavity flange and the welded area of a cavity body with the flange. This analysis is also necessary to predict a potential deformation of the cavity under the combined force of atmospheric pressure and the seating load of the seal. Modeling of this assembly has been achieved using both HKS/Abaqus and COSMOS

Vacuum seals design and testing for a linear accelerator of the National Spallation Neutron Source

International Nuclear Information System (INIS)

Chen, Z.; Gautier, C.; Hemez, F.; Bultman, N.K.

2000-01-01

Vacuum seals are very important to ensure that the Spallation Neutron Source (SNS) Linac has an optimum vacuum system. The vacuum joints between flanges must have reliable seals to minimize the leak rate and meet vacuum and electrical requirements. In addition, it is desirable to simplify the installation and thereby also simplify the maintenance required. This report summarizes an investigation of the metal vacuum seals that include the metal C-seal, Energized Spring seal, Helcoflex Copper Delta seal, Aluminum Delta seal, delta seal with limiting ring, and the prototype of the copper diamond seals. The report also contains the material certifications, design, finite element analysis, and testing for all of these seals. It is a valuable reference for any vacuum system design. To evaluate the suitability of several types of metal seals for use in the SNS Linac and to determine the torque applied on the bolts, a series of vacuum leak rate tests on the metal seals have been completed at Los Alamos Laboratory. A copper plated flange, using the same type of delta seal that was used for testing with the stainless steel flange, has also been studied and tested. A vacuum seal is desired that requires significantly less loading than a standard ConFlat flange with a copper gasket for the coupling cavity assembly. To save the intersegment space the authors use thinner flanges in the design. The leak rate of the thin ConFlat flange with a copper gasket is a baseline for the vacuum test on all seals and thin flanges. A finite element analysis of a long coupling cavity flange with a copper delta seal has been performed in order to confirm the design of the long coupling cavity flange and the welded area of a cavity body with the flange. This analysis is also necessary to predict a potential deformation of the cavity under the combined force of atmospheric pressure and the seating load of the seal. Modeling of this assembly has been achieved using both HKS/Abaqus and COSMOS

Status of LMFBR development project in Japan

International Nuclear Information System (INIS)

Nagane, G.; Akebi, M.; Matsuno, Y.

1987-01-01

Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

Sodium fire studies in France safety tests and applications on an LMFBR

International Nuclear Information System (INIS)

Fruchard, Y.; Colome, J.; Malet, J.C.; Berlin, M.; de Cuy, G.D.; Justin, J.; Duco, J.; Fourest, B.

1976-01-01

The risk of sodium fires in an LMFBR requires thorough analysis, and the possible consequences of an accidental fire must be accurately determined. Not only must means of prevention and detection be perfected, but techniques must be developed to limit the damage caused by a fire: extinguishment, aerosol containment, protection of reactor structures. The program currently undertaken by the CEA's Nuclear Safety Department covering these problems is described. The major results obtained as well as their application to the SUPER-PHENIX reactor are included

Seal-rotordynamic-coefficient Test Results for a Model SSME ATD-HPFTP Turbine Interstage Seal with and Without a Swirl Brake

Science.gov (United States)

Childs, Dara W.; Ramsey, Christopher

1991-01-01

The predictions of Scharrer's (1988) theory for rotordynamic coefficients of labyrinth gas seals were compared with measurements for a model SSME Alternate Turbopump Development High Pressure Fuel Turbopump with and without swirl brakes. Using the test apparatus described by Childs et al., tests were conducted with supply pressures up to 18.3 bars and speeds up to 16,000 rpm. Seal back pressure was controlled to provide four pressure ratios at all supply pressures. No measurable differences in leakage was detected for the seal with and without the swirl brakes. Comparisons of the measurement results for the seal without a swirl brake with the Scharrer theory showed that the theory can be used only to provide design guidelines; systematic differences were observed between theory and experiment due to changes in running speed, supply pressure, and pressure ratio.

Seal-rotordynamic-coefficient test results for a model SSME ATD-HPFTP turbine interstate seal with and without a swirl brake

Science.gov (United States)

Childs, D. W.; Ramsey, C.

1991-01-01

The predictions of Scharrer's (1988) theory for rotordynamic coefficients of labyrinth gas seals were compared with measurements for a model SSME Alternate Turbopump Development High-Pressure Fuel Turbopump with and without swirl brakes. Using the test apparatus described by Childs et al. (1986, 1990), tests were conducted with supply pressures up to 18.3 bars and speeds up to 16,000 rpm. Seal back pressure was controlled to provide four pressure ratios at all supply pressures. No measurable difference in leakage was detected for the seal with and without the swirl brakes. Comparisons of the measurement results for the seal without a swirl brake with the Scharrer theory showed that the theory can be used only to provide design guidelines; systematic differences were observed between theory and experiment due to changes in running speed, supply pressure, and pressure ratio.

Thermal hydraulic analysis of the FFTF core using SUPERENERGY-2

International Nuclear Information System (INIS)

Cramer, E.R.; Basehore, K.L.

1980-01-01

SUPERENERGY-2 is the latest steady-state code in the ENERGY series, combining all of the desirable features of the previous ENERGY-I and SUPERENERGY versions in an optimized form. The result is an easily redimensionable, multiassembly code with many user-convenience features, such as automatic noding and a default constitutive package, that help minimize the effort and time associated with setting up large forced-convection problems. Improvements in physical modeling include generalized facial boundary conditions, duct wall gamma heating, and a model for double-ducted assemblies. The latter is used for modeling both multiduct test and absorber assemblies. SUPERENERGY-2 was used to calculate the temperature distribution in the first six rows of the FFTF core

Leakage test evaluation used for qualification of iodine-125 seeds sealing

International Nuclear Information System (INIS)

Feher, Anselmo; Rostelato, Maria E.C.M.; Zeituni, Carlos A.; Calvo, Wilson A.P.; Somessari, Samir L.; Moura, Joao A.; Moura, Eduardo S.; Souza, Carla D.; Rela, Paulo R.

2009-01-01

The prostate cancer is a problem of public health in Brazil, and the second cause of cancer deaths in men, exceeded only by lung cancer. Among the possible treatments available for prostate cancer is brachytherapy, in which small seeds containing Iodine-125 radioisotope are implanted in the prostate. The seed consists of a sealed titanium tube measuring 0.8 mm external diameter and 4.5 mm in length, containing a central silver wire with adsorbed Iodine-125. The tube sealing is made with titanium at the ends, using electric arc welding or laser process. This sealing must be leakage-resistant and free of cracks, therefore avoiding the Iodine-125 to deposit in the silver wire to escape and spread into the human body. To ensure this problem does not occur, rigorous leakage tests, in accordance with the standard Radiation protection - Sealed Radioactive Sources - leakage Test Methods - ISO 9978, should be applied. The aim of this study is to determine, implement and evaluate the leakage test to be used in the Iodine-125 seeds production, in order to qualify the sealing procedure. The standard ISO 9978 presents a list of tests to be carried out according to the type of source. The preferential methods for brachytherapy sources are soaking and helium. To assess the seeds leakage, the method of immersion test at room temperature was applied. The seeds are considered leakage-free if the detected activity does not exceed the 185 Bq (5 nCi). An Iodine standard was prepared and its value determined in a sodium iodide detector. A liquid scintillation counter was calibrated with the standard for seeds leakage tests. Forty-eight seeds were welded for these tests. (author)

Refractory metal carbide coatings for LMFBR applications: a systems approach

International Nuclear Information System (INIS)

Gotschall, H.L.; Ople, F.S.; Riccardella, P.C.

1975-01-01

The selection, testing and improvement of high density, tightly bonded plasma and detonation gun coatings designed to meet LMFBR core component criteria are described. The process descriptions include a review of the important developments in substrate surface preparation which were required to ensure strong bonding and to minimize interface contamination. Coating finishing techniques which were developed to optimize friction behavior are also described

Testing of molded high temperature plastic actuator road seals for use in advanced aircraft hydraulic systems

Science.gov (United States)

Waterman, A. W.; Huxford, R. L.; Nelson, W. G.

1976-01-01

Molded high temperature plastic first and second stage rod seal elements were evaluated in seal assemblies to determine performance characteristics. These characteristics were compared with the performance of machined seal elements. The 6.35 cm second stage Chevron seal assembly was tested using molded Chevrons fabricated from five molding materials. Impulse screening tests conducted over a range of 311 K to 478 K revealed thermal setting deficiencies in the aromatic polyimide molding materials. Seal elements fabricated from aromatic copolyester materials structurally failed during impulse cycle calibration. Endurance testing of 3.85 million cycles at 450 K using MIL-H-83283 fluid showed poorer seal performance with the unfilled aromatic polyimide material than had been attained with seals machined from Vespel SP-21 material. The 6.35 cm first stage step-cut compression loaded seal ring fabricated from copolyester injection molding material failed structurally during impulse cycle calibration. Molding of complex shape rod seals was shown to be a potentially controllable technique, but additional molding material property testing is recommended.

Fast Flux Test Facility sodium pump operating experience - mechanical

International Nuclear Information System (INIS)

Buonamici, R.

1987-11-01

The Heat Transport System (HTS) pumps were designed, fabricated, tested, and installed in the Fast Flux Test Facility (FFTF) Plant during the period from September 1970 through July 1977. Since completion of the installation and sodium fill in December 1978, the FFTF Plant pumps have undergone extensive testing and operation with HTS testing and reactor operation. Steady-state hydraulic and mechanical performances have been and are excellent. In all, FFTF primary and secondary pumps have operated in sodium for approximately 75,000 hours and 79,000 hours, respectively, to August 24, 1987

Attenuation of airborne debris from LMFBR accidents

International Nuclear Information System (INIS)

Morewitz, H.A.; Johnson, R.P.; Nelson, C.T.; Vaughan, E.U.; Guderjahn, C.A.; Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

1978-01-01

Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m 3 closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approximately 1,000 g/m 3 ), high turbulence, and high temperature (approximately 2000 0 C); and aerosol transport through various leak paths. These studies have shown that tittle, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building. (author)

Chemical factors affecting fission product transport in severe LMFBR accidents

International Nuclear Information System (INIS)

Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

1984-10-01

This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

Neutronic characteristics simulation of LMFBR of great size

International Nuclear Information System (INIS)

Kim, Y.C.

1987-09-01

The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr

Guidelines for testing sealed radiation sources

International Nuclear Information System (INIS)

1989-01-01

These guidelines are based on article 16(1) of the Ordinance on the Implementation of Atomic Safety and Radiation Protection dated 11 October 1984 (VOAS), in connection with article 36 of the Executory Provision to the VOAS, of 11 October 1984. They apply to the testing of sealed sources to verify their intactness, tightness and non-contamination as well as observance of their fixed service time. The type, scope and intervals of testing as well as the evaluation of test results are determined. These guidelines also apply to the testing of radiation sources forming part of radiation equipment, unless otherwise provided for in the type license or permit. These guidelines enter into force on 1 January 1990

Measurements of dynamic shape factors of LMFBR aggregate aerosols

International Nuclear Information System (INIS)

Allen, M.D.; Moss, O.R.; Briant, J.K.

1980-01-01

Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

Accident considerations in LMFBR design

International Nuclear Information System (INIS)

Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

1975-12-01

LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

Sealing arrangement for radioactive material

International Nuclear Information System (INIS)

Gray, I.L.S.; Sievwright, R.W.T.; Elliott, J.C.

1993-01-01

A sealing arrangement for hermetically sealing two mating surfaces comprises two seals arranged to lie between the surfaces. Each seal provides hermetic sealing over a respective different temperature range and lie serially along the surfaces between the regions to be isolated. A main seal integrity test arrangement is provided in the form of a port and passage. This allows for the introduction of a fluid into or the evacuation of a region between the two seals to detect a leak. The port is also provided with at least two test port seals which seal with a plug. The plug is also provided with a test port to allow the integrity of the test port seal to be tested. (UK)

The Solid Waste Transfer Pit an FFTF [Fast Flux Test Facility] station for fuel and reflector transfer

International Nuclear Information System (INIS)

Milewski, A.J.; Thomson, J.D.

1985-01-01

This paper describes the design aspects of a proposed FFTF facility called the Solid Waste Transfer Pit (SWTP). This new facility will be used as a transfer station for reflectors and/or low decay heat fuel assemblies when movement from the Fuel Storage Facility for sodium removal and subsequent reprocessing is to be accomplished. The SWTP utilizes a shielded cell cover with an integral manually-operated gate valve for top loading of a transfer vessel. The cost-effective design discussed herein provides a practical approach using state-of-the-art concepts while assuring safe and reliable operation

LMFBR: safety aspects

International Nuclear Information System (INIS)

Natta, M.

1990-01-01

This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

Safety consequences of local initiating events in an LMFBR

Energy Technology Data Exchange (ETDEWEB)

Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

1975-12-01

The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

Safety consequences of local initiating events in an LMFBR

International Nuclear Information System (INIS)

Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

1975-12-01

The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered

Final report of the borehole, shaft, and tunnel sealing test. Vol. 3

International Nuclear Information System (INIS)

Pusch, R.; Boergesson, L.; Ramqvist, G.

1987-02-01

Like the borehole and shaft plugging tests, the tunnel test gave evidence of the very effective sealing power of Na betonite. The test arrangement consisted of a 9 m long 1.5 m diameter steel tube surrounded by sand and cast in concrete plugs at each end. These plugs contained bentonite forming 'O-ring' sealings at the concrete/rock interface. The test had the form of injecting water into the sand and measuring the leakage that took place through the adjacent rock and along the plug. It was concluded that the drop in leakage from more than 200 l at 100 kPa water pressure early in the test to 75 l at 3 MPa pressure at the end was due partly to the swelling pressure exerted by the bentonite on the rock and by penetration of bentonite into water-bearing rock fractures. The major sealing process appears to be the establishment of a very tight bentonite/rock interface. (orig./HP)

Cover-gas-seal component development: dynamic inflatable-plug seal improvement

International Nuclear Information System (INIS)

Horton, P.H.

1977-01-01

This report documents the 1) radial compliance and 2) low friction coating tests conducted on the CRBRP Rotating Plug Inflatable Seals per test plan N707TR810014. Test results show that narrowing the seal blade from 0.25 to 0.12 in. will effectively reduce dynamic drag from 30 to 20 lb/ft under nominal conditions and will increase seal radial compliance from 0.12 to 0.30 in. without an unacceptable rise in dynamic drag. Tests also demonstrated that application of a teflon coating to the seal wear surface reduced breakaway drag by 25% based on results of comparison dwells

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

International Nuclear Information System (INIS)

Kirk, W.L.

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

Energy Technology Data Exchange (ETDEWEB)

Kirk, W.L. (comp.)

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

Thermal-buckling analysis of an LMFBR overflow vessel

International Nuclear Information System (INIS)

Severud, L.K.

1983-01-01

During a reactor scram, cold sodium flows into the hot overflow vessel. The effect on the vessel is a compressive thermal stress in a zone just above the sodium level. This condition must be sufficiently controlled to preclude thermal buckling. Also, under repeated scrams, the vessel should not suffer thermal stress low cycle fatigue. To evaluate the closeness to buckling and satisfaction of ASMA Code limits, a combination of simple approximations, detailed elastic shell buckling analyses, and correlations to results of thermal buckling tests were employed. This paper describes the analysis methods, special considerations, and evaluations accomplished for this FFTF vessel to assure satisfaction of ASME buckling design criteria, rules, and limits

Research report on design allowable values of structural materials for LMFBR

International Nuclear Information System (INIS)

1978-11-01

The present report is composed of following two main parts. i) review and re-evaluation on test results by FCI Sub-committee studies, performed from 1973 to 1976, ii) review on procedures for determining design allowable values of structural materials for LMFBR components. Re-evaluation works have been made on monotonic tensile properties at elevated temperatures, creep and creep rupture properties, creep-fatigue properties (strain rate and tensile strain hold time effects on strain fatigue properties at elevated temperatures) of Types 316 and 304 stainless steel and 2 1/4Cr-1Mo steel (base and weld metals) produced in Japan. In the first half of the present report, creep-fatigue test results obtained by FCI Sub-committee studies are subjected to re-evaluation by the present P-FCI Sub-committee. Reviews have been made on testing methods on FCI's-creep-fatigue experiments with other test data of the test materials; high temperature monotonic tensile data, creep and creep rupture data, and origin of the test materials. The data of FCI studies are compared with other reference data obtained by several Japanese laboratories. In the latter half of the present report, procedures including ASME's are reviewed for setting design allowable values for LMFBR components on the basis of high temperature strength properties obtained with materials produced in Japan. A creep rupture data of Japanese steels are issued and examined to make proposal for a design allowable stress of S sub(t) through parameter survey. (author)

Regulatory concerns for leakage testing of packagings with three O-ring closure seals

International Nuclear Information System (INIS)

Oras, J.J.; Towell, R.H.; Wangler, M.E.

1997-01-01

The American National Standard for Radioactive Materials--Leakage Tests on Packages for Shipment (ANSI N14.5) provides guidance for leakage rate testing to show that a particular packaging complies with regulatory requirements and also provides guidance in determining appropriate acceptance criteria. Recent radioactive packagings designs have incorporated three O-ring closure seals, the middle O-ring being the containment seal. These designs have the potential for false positive results of leakage rate tests. The volume between the containment O-ring and the inner O-ring is used for the helium gas required for the leakage rate tests to reduce both the amount of helium used and the time required to conduct the tests. A leak detector samples the evacuated volume between the outer O-ring and the containment O-ring. False positive results can be caused in two ways, a large leakage in the containment seal or leakage in the inner seal. This paper will describe the problem together with possible solutions/areas that need to be addressed in a Safety Analysis Report for Packagings before a particular packaging design can be certified for transport

Utilizing the Fast Flux Test Facility for international passive safety testing

International Nuclear Information System (INIS)

Shen, P.K.; Padilla, A.; Lucoff, D.M.; Waltar, A.E.

1991-01-01

A two-phased approach has been undertaken in the Fast Flux Test Facility (FFTF) to conduct passive safety testing. Phase I (1986 to 1987) was structured to obtain an initial understanding of the reactivity feedback components. The planned Phase II (1992 to 1993) international program will extend the testing to include static and dynamic feedback measurements, transient and demonstration tests, and gas expansion module (GEM) reactivity tests. The primary objective is to meet the needs for safety analysis code validation, with particular emphasis on reducing the uncertainties associated with structure reactivity feedback. Program scope and predicted FFTF responses are discussed and illustrated. (author)

Airborne effluent control for LMFBR fuel reprocessing plants

International Nuclear Information System (INIS)

Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

1976-01-01

A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

Hydrogen jet recombination under postulated LMFBR accident conditions

International Nuclear Information System (INIS)

Wierman, R.W.

1977-01-01

Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

Inelastic finite element cyclic analysis of a nozzle-to-cylinder intersection

International Nuclear Information System (INIS)

Barsoum, R.S.; Loomis, R.W.

1976-01-01

A finite element elastic-plastic and creep analysis of a nozzle-to-cylinder intersection subject to cyclic thermal shock, internal pressure, and mechanical loads is presented. The nozzle configuration is that of the intermediate heat exchanger (IHX) for the Fast Flux Test Facility (FFTF). The analysis was performed using the general purpose program MARC. Both the elastic and inelastic results of the analysis are presented. The intention of this study to analytically investigate the applicability of simplified ratchetting and creep-fatigue rules for LMFBR components, as a part of a program covering various geometries and loadings

Labyrinth Seal Flutter Analysis and Test Validation in Support of Robust Rocket Engine Design

Science.gov (United States)

El-Aini, Yehia; Park, John; Frady, Greg; Nesman, Tom

2010-01-01

High energy-density turbomachines, like the SSME turbopumps, utilize labyrinth seals, also referred to as knife-edge seals, to control leakage flow. The pressure drop for such seals is order of magnitude higher than comparable jet engine seals. This is aggravated by the requirement of tight clearances resulting in possible unfavorable fluid-structure interaction of the seal system (seal flutter). To demonstrate these characteristics, a benchmark case of a High Pressure Oxygen Turbopump (HPOTP) outlet Labyrinth seal was studied in detail. First, an analytical assessment of the seal stability was conducted using a Pratt & Whitney legacy seal flutter code. Sensitivity parameters including pressure drop, rotor-to-stator running clearances and cavity volumes were examined and modeling strategies established. Second, a concurrent experimental investigation was undertaken to validate the stability of the seal at the equivalent operating conditions of the pump. Actual pump hardware was used to construct the test rig, also referred to as the (Flutter Rig). The flutter rig did not include rotational effects or temperature. However, the use of Hydrogen gas at high inlet pressure provided good representation of the critical parameters affecting flutter especially the speed of sound. The flutter code predictions showed consistent trends in good agreement with the experimental data. The rig test program produced a stability threshold empirical parameter that separated operation with and without flutter. This empirical parameter was used to establish the seal build clearances to avoid flutter while providing the required cooling flow metering. The calibrated flutter code along with the empirical flutter parameter was used to redesign the baseline seal resulting in a flutter-free robust configuration. Provisions for incorporation of mechanical damping devices were introduced in the redesigned seal to ensure added robustness

Design of a single variable helium effects experiment for irradiation in FFTF [Fast Flux Test Facility] using alloys enriched in nickel 59

International Nuclear Information System (INIS)

Simons, R.L.; Brager, H.R.; Matsumoto, W.Y.

1986-03-01

Nickel enriched in nickel 59 was extracted from the fragments of a fracture toughness specimen of Inconel 600 irradiated in the Engineering Test Reactor (ETR). The nickel contained 2.0% nickel 59. Three heats of austenitic steel doped with nickel-59 were prepared and inserted in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The experiment was single variable in helium effects because chemically identical alloys without nickel-59 were being irradiated side by side with the doped material. The alloys doped with nickel 59 produced 10 to 100 times more helium than the control alloys. The materials included ternary and quaternary alloys in the form of transmission electron microscope (TEM) discs and miniature tensile specimens. The helium to dpa ratio was in the range 5 to 35 and was nearly constant throughout the irradiation. The exposures ranged from 0.25 to 50 displacements per atom (dpa) over the duration of the experiment. The irradiation temperatures covered the range of 360 to 600 0 C

Preventive maintenance and load testing of fixed position cranes in support of major operations

International Nuclear Information System (INIS)

Detrick, C.K.

1980-01-01

This paper will address load testing and preventive maintenance of fixed in-place cranes in general, and maintenance and load testing of the 200 ton Polar Gantry Crane at the FFTF in particular. This paper also covers the installation of a 100-ton bridge crane in the FFTF's Reactor Service Building, as well as use of these cranes in making important lifts of FFTF equipment

Fast Flux Test Facility project plan. Revision 2

Energy Technology Data Exchange (ETDEWEB)

Hulvey, R.K.

1995-11-01

The Fast Flux Test Facility (FFTF) Transition Project Plan, Revision 2, provides changes to the major elements and project baseline for the deactivation activities necessary to transition the FFTF to a radiologically and industrially safe shutdown condition.

Fast Flux Test Facility project plan. Revision 2

International Nuclear Information System (INIS)

Hulvey, R.K.

1995-11-01

The Fast Flux Test Facility (FFTF) Transition Project Plan, Revision 2, provides changes to the major elements and project baseline for the deactivation activities necessary to transition the FFTF to a radiologically and industrially safe shutdown condition

An Investigation of Loop Seal Clearings in ATLAS SBLOCA Tests

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeonsik; Cho, Seok; Kang, Kyoungho; Park, Hyunsik; Min, Kyeongho; Choi, Namhyeon; Park, Jonggook; Kim, Bokdeuk; Choi, Kiyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

In most of the SBLOCA cases, the pressure of the upper-head region will increase mainly owing to the accumulated steam and water inventory in the upper-plenum. This build-up pressure acts as a suppression force to the core water level, and resultantly the core water level will decrease possibly up to and/or below the top of the active core region. Simultaneously, the downcomer water level will increase owing to the evacuated water inventory from the lower part of the core region. This unbalanced hydro-static pressure between the core and downcomer region acts as a potential pushing force to the reactor coolant pump (RCP) side intermediate leg. The potential pushing force will be increased with time to overcome the hydro-static head in the upflow intermediate leg. The unbalanced hydro-static pressure can finally be dissolved with the occurrence of the loop seal clearing. A minimum core collapsed water level, located below the elevation of the loop seal bottom leg in the ATLAS tests, is taken at this time. Since the loop seal bottom leg is located below the core top for typical PWR plants such as an APR1400, the water level depression may uncover the core upper regions until the core water level recovers with the progress of the clearing of the loop seal upflow leg. At this moment, the core temperature may increase to a peak cladding temperature (PCT) owing to an excessive core uncovery by the minimum core collapsed water level. Therefore, the loop seal clearing phenomenon is very important with respect to the PCT occurrence, which is one of the most important parameters to insure the safety of the reactor system. The loop seal clearing behavior seems to be closely related to the break location and break size. Usually, a loop seal in the break loop is cleared first, and the number of loop seal clearings is dependent on the break size. The larger the break size, the more the loop seals that are cleared. An investigation of LSC in the SBLOCA for DVI line and CL breaks

Mechanical seals

CERN Document Server

Mayer, E

1977-01-01

Mechanical Seals, Third Edition is a source of practical information on the design and use of mechanical seals. Topics range from design fundamentals and test rigs to leakage, wear, friction and power, reliability, and special designs. This text is comprised of nine chapters; the first of which gives a general overview of seals, including various types of seals and their applications. Attention then turns to the fundamentals of seal design, with emphasis on six requirements that must be considered: sealing effectiveness, length of life, reliability, power consumption, space requirements, and c

Oil in the FFTF secondary loop cover gas piping. Final unusual occurrence report

International Nuclear Information System (INIS)

Kuechle, J.D.

1981-01-01

The final unusual occurrence report describes the discovery of oil in the FFTF secondary sodium system cover gas piping. A thorough evaluation has been performed and corrective actions have been implemented to prevent a recurrence of this event

Development of an 85,000 gpm (19,303 m3/h) LMFBR primary pump

International Nuclear Information System (INIS)

Zerinvary, M.C.; Wagner, E.W.

1984-01-01

The development of an 85,000 gpm two-stage primary pump for liquid metal fast breeder reactor (LMFBR) applications is described. The design was supported by air and cavitation model testing of the hyraulics, and development and feature testing of the level control system and the adjustable frequency solid state power supply. Important fabrication and water test items are also discussed, along with some unique assembly tooling requirements

Field Test of Advanced Duct-Sealing Technologies Within the Weatherization Assistance Program

Energy Technology Data Exchange (ETDEWEB)

Ternes, MP

2001-12-05

A field test of an aerosol-spray duct-sealing technology and a conventional, best-practice approach was performed in 80 homes to determine the efficacy and programmatic needs of the duct-sealing technologies as applied in the U.S. Department of Energy Weatherization Assistance Program. The field test was performed in five states: Iowa, Virginia, Washington, West Virginia, and Wyoming. The study found that, compared with the best-practice approach, the aerosol-spray technology is 50% more effective at sealing duct leaks and can potentially reduce labor time and costs for duct sealing by 70%, or almost 4 crew-hours. Further study to encourage and promote use of the aerosol-spray technology within the Weatherization Assistance Program is recommended. A pilot test of full production weatherization programs using the aerosol-spray technology is recommended to develop approaches for integrating this technology with other energy conservation measures and minimizing impacts on weatherization agency logistics. In order to allow or improve adoption of the aerosol spray technology within the Weatherization Assistance Program, issues must be addressed concerning equipment costs, use of the technology under franchise arrangements with Aeroseal, Inc. (the holders of an exclusive license to use this technology), software used to control the equipment, safety, and training. Application testing of the aerosol-spray technology in mobile homes is also recommended.

CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

International Nuclear Information System (INIS)

D.R. Moscalu

2002-01-01

The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M andO 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M andO 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k eff ) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-0000101 (BSC 2002)

Compendium of computer codes for the safety analysis of LMFBR's

International Nuclear Information System (INIS)

1975-06-01

A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

Applications of simulation experiments in LMFBR core materials technology

International Nuclear Information System (INIS)

Appleby, W.K.

1976-01-01

The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

A review on the welding technology for the sealing of irradiation test fuel element

Energy Technology Data Exchange (ETDEWEB)

Lee, J. W.; Kang, Y. H.; Kim, B. G.; Joo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

2000-02-01

For the irradiation test of nuclear fuel in a research reactor, the fuel manufacturing technology should be developed in advance. Highly radioactive fission products are produced and can be released from the fuel materials during irradiation. Therefore, The sealing of the test is one of the most important procedure among the test fuel manufacturing processes, considering its impacts on the safety of a reactor operation.many welding techniques such as TIG, EBW, LBW, upset butt welding and flash welding are applied in sealing the end of fuel elements. These welding techniques are adopted in conjunction with the weld material, weldability, weld joint design and cost effectiveness. For fuel irradiation test, the centerline temperature of fuel pellets is one of the important item to be measured. For this, a thermocouple is installed into the center of the fuel pellet. The sealing of the penetration hole of the thermocouple sheath should be conducted and the hole should be perfectly sealed using the dissimilar metal joining technique. For this purpose, the dissimilar metal welding between zircaloy-4 and Inconel or stainless steel is needed to be developed. This report describes the techniques sealing the end cap and the penetration of a thermocouple sheath by welding. (author)

The large scale in-situ PRACLAY heater and seal tests in URL HADES, Mol, Belgium

Energy Technology Data Exchange (ETDEWEB)

Xiangling Li; Guangjing Chen; Verstricht, Jan; Van Marcke, Philippe; Troullinos, Ioannis [ESV EURIDICE, Mol (Belgium)

2013-07-01

In Belgium, the URL HADES was constructed in the Boom Clay formation at the Mol site to investigate the feasibility of geological disposal in a clay formation. Since 1995, the URL R and D programme has focused on large scale demonstration tests like the PRACLAY Heater and Seal tests. The main objective of the Heater Test is to demonstrate that the thermal load generated by the heat-emitting waste will not jeopardise the safety functions of the host rock. The primary objective of the Seal Test is to provide suitable hydraulic boundary conditions for the Heater Test. The Seal Test also provides an opportunity to investigate the in-situ behaviour of a bentonite-based EBS. The PRACLAY gallery was constructed in 2007 and the hydraulic seal was installed in 2010. The bentonite is hydrated both naturally and artificially. The swelling, total pressure and pore pressure of the bentonite are continuously measured and analysed by numerical simulations to get a better understanding of this hydration processes. The timing of switching on the heater depends on the progress of the bentonite hydration, as a sufficient seal swelling is needed to fulfill its role. A set of conditions to be met for the heater switch-on and its schedule will be given. (authors)

COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

International Nuclear Information System (INIS)

Rhodes, C.A.

1984-12-01

This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

Pump, sodium, inducer, intermediate size (ISIP) (impeller/inducer/diffuser retrofit)

International Nuclear Information System (INIS)

Paradise, D.R.

1978-01-01

This specification defines the requirements for the Intermediate-Size Inducer Pump (ISIP), which is to be made by replacing the impeller of the FFTF Prototype Pump with a new inducer, impeller, diffuser, seal, and necessary adapter hardware. Subsequent testing requirements of the complete pump assembly are included

Overview of current activities relevant to structural analysis on LMFBR in Japan

International Nuclear Information System (INIS)

Ichimiya, Masakazu

1983-01-01

This paper presents the structural analysis activities on LMFBR in Japan. The structural analysis activities on LMFBR in Japan have been made mainly toward the validation of the rules of high temperature structural design guide which is to be used for the design of Class 1 components for elevated temperature service of the prototype fast breeder reactor, Monju. Main features of these analyses are as follows. (1) Since the design by elastic analysis is intended in the high temperature structural design guide of Monju, a large progress has been made in the bounding technique for high temperature inelastic behaviors, particularly the elastic follow-up. (2) There has been a progress in the clarification of the creep behavior in order to evaluate creep damage adequately. (3) Analysis techniques and design rules for piping have been developed with considerable emphasis. In addition, buckling analyses were performed considering the thin structures with low internal pressure in Monju components. Further test and analysis were made on ratcheting. (author)

PNC status report on leak detector development for LMFBR steam generators

International Nuclear Information System (INIS)

Kuroha, M.; Sato, M.

1984-01-01

Chemical and acoustic type leak detectors have been developed for detecting a small sodium-water reaction in an LMFBR steam generator. This paper presents a summary of the development. (1) Test results on PNC type in-sodium hydrogen meters including a description of the structure, the long-term reliability and the durability, and the improved meter with an orifice, (2) Development of in-cover gas hydrogen meters, (3) Hydrogen detection tests and analyses, (4) Operating experiences of electrochemical in-sodium oxygen meters, and (5) Basic studies on acoustic characteristics of the sodium-water reaction. (author)

Fretting and wear of stainless and ferritic steels in LMFBR steam generators

International Nuclear Information System (INIS)

Lewis, M.W.J.; Campbell, C.S.

1981-01-01

Steam generators for LMFBR's may be subject to both fretting wear as a result of flow-induced vibrations and to wear from larger amplitude sliding movements from thermal changes. Results of tests simulating the latter are given for stainless and ferritic steels. For the assessment of fretting wear damage, vibration assessments must be combined with data on specific wear rates. Test mechanisms used to study fretting in sodium covering impact, impact-slide and pure rubbing are described and results presented. (author)

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

Methods used for analysis of material behaviour, accident phenomenology and integrated accident calculations are reviewed. Applications of these methods to hypothetical LOF and TOP accidents are discussed. Recent results obtained from applications to FFTF and CRBRP are presented. (author)

LMFBR Blanket Physics Project progress report No. 2

International Nuclear Information System (INIS)

Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

1971-01-01

This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

Performance measurements at the fast flux test facility

International Nuclear Information System (INIS)

Baumhardt, R.J.; Newland, D.J.; Praetorius, P.R.

1987-01-01

In 1984, Fast Flux Test Facility (FFTF) management recognized the need to develop a measurement system that would quantify the operational performance of the FFTF and the human resources needed to operate it. Driven by declining budgets and the need to safely manage a manpower rampdown at FFTF, an early warning system was developed. Although the initiating event for the early warning system was the need to safely manage a manpower rampdown, many related uses have evolved. The initial desired objective for the FFTF performance measurements was to ensure safety and control of key performance trends. However, the early warning system has provided a more quantitative, supportable basis upon which to make decisions. From this initial narrow focus, efforts in the FFTF plant and supporting organizations are leading to measurement of and, subsequently, improvements in productivity. Pilot projects utilizing statistical process control have started with longer range productivity improvement

Mechanical seal program

International Nuclear Information System (INIS)

Lowery, G.B.

1983-01-01

The experimental plans and timing for completion of the mechanical seal program for both the slurry and transfer pumps are given. The slurry pump seal program will be completed by April 1984 with turnover of two seals in pumps to SRP Tank 15H. Transfer pump seal design will be released for plant use by May 1984. Also included are various other pump and seal related tests

Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

International Nuclear Information System (INIS)

Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

1976-01-01

Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

Structural analysis for LMFBR applications

International Nuclear Information System (INIS)

1983-01-01

Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start

Structural analysis for LMFBR applications

International Nuclear Information System (INIS)

Vaze, M.K.K.

1983-01-01

The use of elastic analysis for structural design of LMFBR components is discussed. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed Prototype Fast Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is same as that of Rapsodie. Nevertheless, the design had to be checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, ASME Code Section III and the Code Case N-47 are used for high temperature design. The problems faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's shakedown and plastic cycling criteria for ratchet free operation to biaxial stress fields

THE LMFBR, key to the future

International Nuclear Information System (INIS)

Chipman, G.L. Jr.

1982-01-01

This survey explains the United States prospects for utilizing the LMFBR as a mean of meeting future energy demands. Nuclear option will represent a good financial investment only when breeder will be proved as a cost-effective option. International cooperation and combined programs are very helpful to develop breeder reactor power resource

Evaluation of methods to leak test sealed radiation sources

International Nuclear Information System (INIS)

Arbeau, N.D.; Scott, C.K.

1987-04-01

The methods for the leak testing of sealed radiation sources were reviewed. One hundred and thirty-one equipment vendors were surveyed to identify commercially available leak test instruments. The equipment is summarized in tabular form by radiation type and detector type for easy reference. The radiation characteristics of the licensed sources were reviewed and summarized in a format that can be used to select the most suitable detection method. A test kit is proposed for use by inspectors when verifying a licensee's test procedures. The general elements of leak test procedures are discussed

Operating conditions of steam generators for LMFBR's

Energy Technology Data Exchange (ETDEWEB)

Ratzel, W

1975-07-01

Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

Operating conditions of steam generators for LMFBR's

International Nuclear Information System (INIS)

Ratzel, W.

1975-01-01

Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

International Nuclear Information System (INIS)

Kryter, R.C.

1975-01-01

Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

A fundamental study on sodium-water reaction in the double-pool-type LMFBR

International Nuclear Information System (INIS)

Yoshida, Kazuo; Akimoto, Tokuzo

1987-01-01

In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

Status of the LMFBR development

Energy Technology Data Exchange (ETDEWEB)

Taylor, J J

1975-01-01

The development of any new power generation system which can make a major contribution to our energy needs is a multi-faceted task involving the utilization of major human and material resources. The LMFBR development, which has the potential for supplying abundant energy for generations, is therefore a large, multi-faceted program. This summary will cover (1) the need for the liquid metal fast breeder reactor, (2) an overall perspective of its development throughout the world, (3) a brief look at the in-depth technological development program in the United States, (4) a description and status of the two major projects now under way in the program, the Fast Flux Test Facility and the Clinch River Breeder Reactor Plant, and (5) a review of the plans for continued development to achieve a reliable, safe and economic power generation system for practical commercial use on the utility networks of the country.

Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

International Nuclear Information System (INIS)

Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

1979-09-01

The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

International Nuclear Information System (INIS)

Henslee, S.P.

1987-03-01

The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

Accelerated test program for sealed nickel-cadmium spacecraft batteries/cells

Science.gov (United States)

Goodman, L. A.

1976-01-01

The feasibility was examined of inducing an accelerated test on sealed Nickel-Cadmium batteries or cells as a tool for spacecraft projects and battery users to determine: (1) the prediction of life capability; (2) a method of evaluating the effect of design and component changes in cells; and (3) a means of reducing time and cost of cell testing.

Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

Energy Technology Data Exchange (ETDEWEB)

Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

1996-10-01

In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

Hermetic Seal Leak Detection Apparatus

Science.gov (United States)

Kelley, Anthony R. (Inventor)

2013-01-01

The present invention is a hermetic seal leak detection apparatus, which can be used to test for hermetic seal leaks in instruments and containers. A vacuum tight chamber is created around the unit being tested to minimize gas space outside of the hermetic seal. A vacuum inducing device is then used to increase the gas chamber volume inside the device, so that a slight vacuum is pulled on the unit being tested. The pressure in the unit being tested will stabilize. If the stabilized pressure reads close to a known good seal calibration, there is not a leak in the seal. If the stabilized pressure reads closer to a known bad seal calibration value, there is a leak in the seal. The speed of the plunger can be varied and by evaluating the resulting pressure change rates and final values, the leak rate/size can be accurately calculated.

Low-Torque Seal Development

Science.gov (United States)

Lattime, Scott B.; Borowski, Richard

2009-01-01

The EcoTurn Class K production prototypes have passed all AAR qualification tests and received conditional approval. The accelerated life test on the second set of seals is in progress. Due to the performance of the first set, no problems are expected.The seal has demonstrated superior performance over the HDL seal in the test lab with virtually zero torque and excellent contamination exclusion and grease retention.

LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

International Nuclear Information System (INIS)

Petrykowski, J.C.; Longest, A.W.

1985-01-01

The transport of uranium dioxide (UO 2 ) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO 2 pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs

A hybrid floating brush seal (HFBS) for improved sealing and wear performance in turbomachinery applications

Science.gov (United States)

Lattime, Scott Byran

A conceptually new type of seal has been developed for gas turbine applications which dramatically reduces wear and leakage associated with current labyrinth and brush seal technologies. The Hybrid Floating Brush Seal (HFBS) combines brush seal and film riding face seal technologies to create a hybrid seal that allows both axial and radial excursions of the sealed shaft, while simultaneously eliminating interface surface speeds (friction and heat) between the rotor and the brush material that characterize standard brush seal technology. A simple test rig was designed to evaluate feasibility of the HFBS under relatively low pressures and rotational speeds (50psig, 5krpm). A second test stand was created to study the effects of centrifugal force on bristle deflection. A third test facility was constructed for prototype development and extensive room temperature testing at moderate pressures and fairly high rotational speeds (100psig, 40krpm). This test rig also allowed the evaluation of the HFBS during axial movement of a rotating shaft. An analytical model to predict the effects of centrifugal force on the bristles of a rotating brush seal was developed. Room temperature analysis of the HFBS proved successful for relatively high operating rotational velocities at moderate pressures with very acceptable leakage rates for gas turbine engines. Brush seals were able to track rotor speeds up to 24krpm while maintaining sealing integrity. The HFBS's ability to function under axial shaft displacement and synchronous dynamic radial loading was also proven successful. Hydrodynamic performance of the face seal was proven to provide adequate stiffness and load carrying capacity to keep the brush seal from contacting the face seal at pressure drops across the brush of up to 100psi. Leakage performance over standard brush seal and labyrinth technology was quite dramatic. The HFBS showed its sealing advantage using much higher radial interference between the rotor and the bristle

Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

International Nuclear Information System (INIS)

Jones, A.V.

1977-01-01

An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

Full-scale demonstration. Fire testing of a system for penetration sealing based on foamed silicone elastomer: Studsvik 77-05-26

International Nuclear Information System (INIS)

Brown, A.

1978-06-01

Testing of a system for making fire retardant penetration seals based on foamed-in-place silicone elastomer is described. The report covers - Concept of fire retardant penetration seals and the Chemtrol system, Design FC 225 - Account of materials used to prepare seals and method of application - Test assembly and full-scale facility at Studsvik - Classification of seals used in demonstration - Diagrams of seals and photographs taken after demonstration

Full-scale demonstration. Fire testing of a system for penetration sealing based on foamed silicone elastomer: Studsvik 77-05-26

International Nuclear Information System (INIS)

Brown, A.

1978-06-01

Testing of a system for making fire retardant penetration seals based on foamed-in-place silicone elastomer is described. The report covers - Concept of fire retardant penetration seals and the Chemtrol system, Design FC 225 - Account of materials used to prepare seals and method of application - Test assembly and full-scale facility at Studsvik - Classification of seals used in demonstration - Diagrams of seals and photographs taken after demonstration (author)

Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

Testing to determine the leakage behavior of inflatable seals subject to severe accident loadings

International Nuclear Information System (INIS)

Parks, M.B.

1988-01-01

Under the sponsorship of the United States Nuclear Regulatory Commission, Sandia National Laboratories is currently developing test validated methods to predict the pressure capacity, at elevated temperatures, of light water reactor (LWR) nuclear containment vessels subject to loads well beyond their design basis - the so-called severe accident. Scale model tests of containments with the major penetrations represented have been carried to functional failure by internal pressurization. Also, combined pressure and elevated temperature tests of typical compression seals and gaskets, a full size personnel airlock, and of typical electrical penetration assemblies (EPAs), have been conducted in order to better understand the leakage behavior of containment penetrations. Because inflatable seals are also a part of the pressure boundary of some containments, it is important to understand their leakage behavior as well. This paper discusses the results of tests that were performed to better define the leakage behavior of inflatable seals when subjected to loads well beyond their design basis

Final report of the borehole, shaft, and tunnel sealing test. Vol.2

International Nuclear Information System (INIS)

Pusch, R.; Boergesson, L.; Ramqvist, G.

1987-01-01

Shaft sealing by use of highly compacted bentonite was investigated in a 14 m long shaft in which two plugs were constructed with a central sand-filled central space for injecting water. A first reference test with concrete plugs was followed by a main test in which the plug material consisted of blocks of highly compacted sodium bentonite powder. In the latter test, the outflow from the injection chamber was only a few percent of that with the concrete plugs, which demonstrates the excellent sealing properties of the clay. The main effect was that practically no water flow took place along the rock/clay interface. The longevity of smectite clay in crystalline rock is sufficient to make bentonite plugs operative for several thousand years. (authors)

Final report of the borehole, shaft and tunnel sealing test. Vol.1

International Nuclear Information System (INIS)

Pusch, R.; Boergesson, L.; Ramqvist, G.

1987-01-01

The Borehole Plugging Experiment comprised field tests of the sealing function and the practicality in handling and application of plugs consisting of segments of perforated metal casings filled with cylindrical blocks of highly compacted sodium bentonite. Preparative tests had shown that the clay swells out through the perforation and embeds the casings. The field tests demonstrated that even very long holes can be effectively sealed by such plugs and that the clay becomes very homogeneous and forms a tight contact with the rock in a relatively short time. By that the plugs become practically impervious and the flow along the clay/rock contact will be insignificant. The longevity of such plugs extends over several thousand years under the conditions that usually prevail in crystalline rock. (authors)

Design and development of face seal type sealing plug for advanced heavy water reactor

International Nuclear Information System (INIS)

Bansal, S.; Bhattacharyya, S.; Patel, R.J.; Agrawal, R.G.; Vaze, K.K.

2005-09-01

Advanced Heavy Water Reactor is a vertical pressure tube type reactor having light water as its coolant and heavy water as moderator. Sealing plug is required to close the pressure boundary of main heat transport system of the reactor by preventing escape of light water/steam From the coolant channel. There are 452 coolant channels in the reactor located in square lattice pitch. Sealing plug is located at the top of each coolant channel (in the top end fitting). Top end fitting is having a stepped bore to create a sealing face. Sealing plug is held through its expanded jaws in a specially provided groove of the end fitting. The plug was designed and prototypes were manufactured considering its functional importance, intricate design and precision machining requirements. Sealing plug consists of about 20 components mostly made up of precipitation hardening stainless steel, which is suitable for water environment and meets other requirements of strength and resistance to wear and galling. Seal disc is a critical component of the sealing plug as it is the pressure-retaining component. It is a circular disc with protruded stem. One face of the seal disc is nickel plated in the peripheral area that creates the sealing by abutting against the sealing face provided in the end fitting. The typical shape and profile of seal disc provides flexibility and allows elastic deformation to assist in locking of sealing plug and creating adequate seating force for effective sealing. Design and development aspects of the sealing plug have been detailed out in this report. Also results of stress analysis and experimental studies for seal disc have been mentioned in the report. Stress analysis and experimental testing was required for the seal disc because high stresses are developed due to its exposure to high pressure and temperature environment of Main Heat Transport system. Hot testing was carried out to simulate the reactor-simulated condition. The performance was found to be

Role of FFTF in assessing structural feedbacks and inherent safety of LMR's

International Nuclear Information System (INIS)

Padilla, A.; Omberg, R.P.; O'Dell, L.D.; Harris, R.A.; Nguyen, D.H.; Waltar, A.E.

1985-03-01

The possibility of developing reactor designs with inherent safety characteristics sufficient to provide ''walk away'' safety is receiving additional emphasis in the LMR program. A key element in this effort is the recognition that LMR's possess safety characteristics above and beyond those employed in past safety review processes. Some of these additional safety characteristics are due to reactivity feedback effects caused by small structural movements during hypothetical severe design transients. The effect of these characteristics upon the behavior of the FFTF under such transients has been assessed and is discussed in this paper. The paper also presents a preliminary test matrix which might allow experimental verification of the structural reactivity feedback effects. Such experimental verification should be very useful to innovative designers seeking to optimize inherent safety. 8 refs., 1 fig., 2 tabs

A probabilistic design method for LMFBR fuel rods

International Nuclear Information System (INIS)

Peck, S.O.; Lovejoy, W.S.

1977-01-01

Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

10 CFR 34.67 - Records of leak testing of sealed sources and devices containing depleted uranium.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Records of leak testing of sealed sources and devices containing depleted uranium. 34.67 Section 34.67 Energy NUCLEAR REGULATORY COMMISSION LICENSES FOR INDUSTRIAL... Requirements § 34.67 Records of leak testing of sealed sources and devices containing depleted uranium. Each...

VAK III. Seals and sealing system

International Nuclear Information System (INIS)

d'Agraives, B.C.; Dal Cero, G.; Debeir, R.; Mascetti, E.; Toornvliet, J.; Volcan, A.

1986-01-01

This report presents the VAK III seals and sealing system, which have been used over a period of two years at the Kahl nuclear facility (Federal Republic of Germany), where field tests and feasibility studies were conducted in order to offer a possible solution for the sealing of LWR fuel assemblies. It has been prepared with the aim of an assessment study to be done at the IAEA. It gives all characteristics and technical descriptions for: the sealing principle, the seal construction, the operating tools, the data processing, the drawings, the publications related to that seal. The main points of progress are: the Strong Random Internal Defects (STRID) incorporated in the seals, allowing the obtention of a good signature stability; the Integrity Check on the Seal Status (broken or not) obtained through a decisive mechanical improvement: the Double Breakage Integrity Check (DOBRIC) and with a better ultrasonic evidence of that status; the provision of new function tools, allowing the performance of Identity Measurements in dry conditions (which means also at the manufacturer plant) or in deeper water (wet storage); the study and development of a new JRC VAK 45 Compact Instrument Box, in which all the measuring functions can be grouped and incorporating an autonomous Minicomputer offering to the Inspection the possibility of performing, on the spot, Correlation and Decision processes. The general benefit of such a feasibility study should be to convince the potential users that such a Safeguards Sealing System can be studied for slightly - or largely - different other applications, provided that the Basic and Operating Functions required to the system be clearly defined, possibly after a common agreement would be stated

Development of Induction Brazing System for Sealing Instrumentation Feed through Part of Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Kim, Kahye; Heo, Sungho; Ahn, Sungho; Joung, Changyoung; Son, Kwangjae; Jung, Yangil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-12-15

To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and 300 .deg. C respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.

Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-10-01

The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors.

Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

International Nuclear Information System (INIS)

1978-10-01

The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

Furnace testing of electrical and pipe-penetration seals based on foamed silicone elastomer: 60, 90, and 120-minute fire ratings

International Nuclear Information System (INIS)

Brown, A.

1979-03-01

Fire tests of foamed silicone seals for electrical and pipe penetrations have been performed using a furnace with temperature control as heat source. The tests were performed in principle in accordance with the requirements of NORDTEST 5A (ISO 834). The purpose of the tests was to obtain appropriate fire ratings for different seal thicknesses. The report covers. - Description of material used to prepare the seals and method of application - Description of furnace test assembly and method of performing test - Listing of penetrating elements and of the thermocouple array used to measure temperature - Curves of thermocouple readouts and photographs of seals during and after completion of the test. (author)

Emergency air cleaning system development for LMFBR containments

International Nuclear Information System (INIS)

McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

1975-01-01

Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

Improved circumferential shaft seal

Science.gov (United States)

Ludwig, L. P.; Strom, T. N.

1974-01-01

Comparative tests of modified and unmodified carbon ring seals showed that addition of helical grooves to conventional segmented carbon ring seals reduced leakage significantly. Modified seal was insensitive to shaft runout and to flooding by lubricant.

Use of the North Sea by harbour seal with special emphasis on the Horns Reef area. Test of prototype GPS/GSM-transmitter on harbour seals in the Sealarium Esbjerg

International Nuclear Information System (INIS)

Tougaard, S.; Tougaard, J.

2004-04-01

A combined GPS-receiver and GSM mobile phone transmitter designed for deployment on wild seals was tested on a captive harbour seal. The unit has been under development in recent years and consists of a GPS-unit for positioning and a GSM mobile phone unit for transmission of logged positions to land. The test took place in the Sealarium at the Fisheries and Maritime Museum, Esbjerg. The unit was glued onto the fur on the head of the seal, where it stayed on for 13 days. Only one position was acquired by the unit and it failed to connect to the GSM-net during the test. Technical information from the constructor of the unit is not available and it is thus not possible to conclude on the reasons why the unit did not function as intended. The behaviour of the seal can however, be ruled out, as close observation and registration of the seals dive behaviour during the first days of deployment showed that the unit was free of the water for sufficiently long periods for positions to be acquired. (au)

FFTF irradiation of fracture mechanics specimens for out-of-core structures

International Nuclear Information System (INIS)

King, D.C.

1978-09-01

The National Program Plan has established data requirements for out-of-core structures for FBRs. Significant FFTF irradiation space with moderate gamma heating levels is required to irradiate relatively large fracture mechanics specimens to total neutron fluences ranging between 5 x 10 21 and 5 x 10 22 n/cm 2 and temperatures which range between 400 0 C (750 0 F) and 650 0 C (1200 0 F). Priority 1 data on stainless steel welds requires a test volume of 7443 cm 3 (454 in 3 ). Priority 2 data on 304 and 316 SS and Inconel 718 materials and Inconel 718 welds requires 2760 cm 3 (168 in 3 ). Priority 3 data on stainless steels, other nickel-base alloys, and ferritics requires 33,118 cm 3 (2021 in 3 ). Priority 4 data at elevated temperatures on stainless steels, other nickel-base alloys and ferritics requires 69,182 cm 3

LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

International Nuclear Information System (INIS)

Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

1979-03-01

Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

Proposed Testing to Assess the Accuracy of Glass-To-Metal Seal Stress Analyses.

Energy Technology Data Exchange (ETDEWEB)

Chambers, Robert S.; Emery, John M; Tandon, Rajan; Antoun, Bonnie R.; Stavig, Mark E.; Newton, Clay S.; Gibson, Cory S; Bencoe, Denise N.

2014-09-01

The material characterization tests conducted on 304L VAR stainless steel and Schott 8061 glass have provided higher fidelity data for calibration of material models used in Glass - T o - Metal (GTM) seal analyses. Specifically, a Thermo - Multi - Linear Elastic Plastic ( thermo - MLEP) material model has be en defined for S S304L and the Simplified Potential Energy Clock nonlinear visc oelastic model has been calibrated for the S8061 glass. To assess the accuracy of finite element stress analyses of GTM seals, a suite of tests are proposed to provide data for comparison to mo del predictions.

Fission-gas bubble modeling for LMFBR accidents

International Nuclear Information System (INIS)

Ostensen, R.W.

1977-01-01

The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

Sodium mists behavior in cover gas space of an LMFBR

International Nuclear Information System (INIS)

Himeno, Y.; Takahashi, J.

1978-03-01

This paper present the sodium mist behavior in Argon cover gas space of an LMFBR experimentally using a test vessel of 1,400 mm in axial length, 305.5 mm in inner diameter and about 100 l in volume. Experiments are consisted with measurements of the mist concentration and the mist gravitational settling flux between the sodium pool temperature range of 290 0 to 520 0 C. The results are discussed under the monosize assumption of the particles, and the particle sizes and evaporation rate are derived. Transient and steady state mist concentration behavior were also investigated. (author)

A review of ANL base technology studies in support of the U.S. LMFBR vibration program

International Nuclear Information System (INIS)

Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Shin, Y.S.

1977-01-01

Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

A review of ANL base technology studies in support of the U.S. LMFBR vibration program

Energy Technology Data Exchange (ETDEWEB)

Wambsganss, M W; Chen, S S [Components Technology Division, Argonne National Laboratory, Argonne, IL (United States); Mulcahy, T M; Shin, Y S

1977-12-01

Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

Stability Analysis of a High-Speed Seal Test Rotor With Marginal and Extended Squeeze-Film Dampers: Theoretical and Experimental Results

Science.gov (United States)

Proctor, Margaret P.; Gunter, Edgar J.

2007-01-01

A case study of a high-speed seal test rotor shows how rotor dynamic analysis can be used to diagnose the source of high vibrations and evaluate a proposed remedy. Experimental results are compared with the synchronous and non-synchronous whirl response analysis of a double overhung, high-speed seal test rotor with ball bearings supported in 5.84- and 12.7-mm-long, un-centered squeeze-film oil dampers. Test performance with the original damper of length 5.84 mm was marginal. Non-synchronous whirling occurred at the overhung seal test disk and there was a high amplitude synchronous response near the drive spline above 32,000 rpm. Nonlinear synchronous unbalance and time transient whirl studies were conducted on the seal test rotor with the original and extended damper lengths. With the original damper design, the nonlinear synchronous response showed that unbalance could cause damper lockup at 33,000 rpm. Alford cross-coupling forces were also included at the overhung seal test disk for the whirl analysis. Sub-synchronous whirling at the seal test disk was observed in the nonlinear time transient analysis. With the extended damper length of 12.7 mm, the sub-synchronous motion was eliminated and the rotor unbalance response was acceptable to 45,000 rpm with moderate rotor unbalance. Seal test rotor orbits and vibration levels with the extended squeeze film dampers showed smooth operation to 40,444 rpm.

Effect of reactor size on the breeding economics of LMFBR blankets

International Nuclear Information System (INIS)

Tagishi, A.; Driscoll, M.J.

1975-02-01

The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

Analysis of ATLAS LTC-04R Test for Loop Seal Reformation Phenomena using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang-Gyu; Kim, Dae-Hun; Kim, Han-Gon [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The loop seal reformation issue was selected to be the analysis topic of the DSP-04 based on the technical discussion between the participants and the operating agencies (KAERI and KINS) and domestic experts meetings. After that, KAERI performed LTC-04R test which is 4 inch top-slot cold-leg break test using ATLAS facility in December 27, 2015. KHNP CRI, as a participant of the DSP-04, performed the blind calculation and open calculation using RELAP5/Mod3.3 patch 3. This paper deals with the results of open calculation for ATLAS LTC-04R test. The results of several sensitivity analyses such as the critical flow modeling sensitivity and break flow system modeling sensitivity will be discussed. Several possible factors in the loop seal reformation behavior are examined in the sensitivity analysis. Heat loss modeling, fine break system modeling, fine loop seal nodalization and off-take modeling are not significant factor in the loop seal reformation. Still critical flow model and discharge coefficient are dominant factors. Based on the ATLAS LTC-04R, Ransom-Trapp model shows better prediction in the break flow than the Henry-Fauske model.

Seals and sealing handbook

CERN Document Server

Flitney, Robert K

2014-01-01

Seals and Sealing Handbook, 6th Edition provides comprehensive coverage of sealing technology, bringing together information on all aspects of this area to enable you to make the right sealing choice. This includes detailed coverage on the seals applicable to static, rotary and reciprocating applications, the best materials to use in your sealing systems, and the legislature and regulations that may impact your sealing choices. Updated in line with current trends this updated reference provides the theory necessary for you to select the most appropriate seals for the job and with its 'Failur

Use of the North Sea by harbour seal with special emphasis on the Horns Reef area. Test of prototype GPS/GSM-transmitter on harbour seals in the Sealarium Esbjerg[Denmark

Energy Technology Data Exchange (ETDEWEB)

Tougaard, S.; Tougaard, J.

2004-04-15

A combined GPS-receiver and GSM mobile phone transmitter designed for deployment on wild seals was tested on a captive harbour seal. The unit has been under development in recent years and consists of a GPS-unit for positioning and a GSM mobile phone unit for transmission of logged positions to land. The test took place in the Sealarium at the Fisheries and Maritime Museum, Esbjerg. The unit was glued onto the fur on the head of the seal, where it stayed on for 13 days. Only one position was acquired by the unit and it failed to connect to the GSM-net during the test. Technical information from the constructor of the unit is not available and it is thus not possible to conclude on the reasons why the unit did not function as intended. The behaviour of the seal can however, be ruled out, as close observation and registration of the seals dive behaviour during the first days of deployment showed that the unit was free of the water for sufficiently long periods for positions to be acquired. (au)

The online sealing performance test of the primary circuit pressure boundary check valve in nuclear power plants

International Nuclear Information System (INIS)

Yang Yunfei; Huang Huimin

2013-01-01

The primary circuit pressure boundary check valves of 320 MW pressurized water reactor is a nuclear grade I key equipment. The sealing demand is very high, which is directly related to the internal leakage rate of the primary circuit system. After the welding check valve is repaired, the sealing performance is judged by color printing checks. If there is water or humid vapor in the pipe, it will affect the accuracy of the color printing checks. For the particularity of the online check valve tightness test, online detecting device is designed by the hydraulic pressure drop method in other nuclear power plants, but the method has some shortcomings and restrictions. In this paper, we design a reliable and portable test equipment by the low-pressure gas seal test flow measurement, which make accurate and quantitative judgment of sealing property after the pressure boundary check valves are repaired. (authors)

LMFBR subassembly response to local pressure loadings: an experimental approach

International Nuclear Information System (INIS)

Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

1975-01-01

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

International Nuclear Information System (INIS)

Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

1978-07-01

The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

Design and economic implications of heterogeneity in an LMFBR core

International Nuclear Information System (INIS)

Orechwa, Y.

1983-01-01

Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

Development and testing of improved polyimide actuator rod seals at higher temperatures for use in advanced aircraft hydraulic systems

Science.gov (United States)

Robinson, E. D.; Waterman, A. W.; Nelson, W. G.

1972-01-01

Polyimide second stage rod seals were evaluated to determine their suitability for application in advanced aircraft systems. The configurations of the seals are described. The conditions of the life cycle tests are provided. It was determined that external rod seal leakage was within prescribed limits and that the seals showed no signs of structural degradation.

Density changes observed in pure molybdenum and Mo-41Re after irradiation in FFTF/MOTA

International Nuclear Information System (INIS)

Garner, F.A.; Greenwood, L.R.

1993-01-01

Pure molybdenum and Mo-41wt% Re, in both the 20% cold-worked and aged and the annealed and aged conditions, were irradiated in FFTF/MOTA to exposures as high as 111 dpa. Pure molybdenum appears to approach a saturation swelling level that is independent of the starting state. Cold-worked and aged molybdenum initially swells at a higher rate than that of solution annealed and aged molybdenum and overshoots the saturation level at lower irradiation temperatures. This requires that part of the accumulated swelling be removed to approach saturation, probably by void shrinkage. The alloy Mo-41Re exhibits a more complex behavior with the annealed and aged condition initially swelling faster, but eventually the density change of both conditions begins to turn downward and tends toward densification. The role of solid transmutation to Tc, Re, and Os is though to be very important in the irradiation behavior of these two metals. Calculations of transmutant generation are provided for FFTF, HFIR and STARFIRE spectra

Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

International Nuclear Information System (INIS)

Moreira, M.L.

1985-01-01

A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

Upgrading inflatable door seals

International Nuclear Information System (INIS)

Sykes, T.M.; Metcalfe, R.; Welch, L.A.; Josefowich, J.M.

1997-01-01

Inflatable door seals are used for airlocks in CANDU stations. They have been a significant source of unreliability and maintenance cost. A program is underway to improve their performance and reliability, backed by environmental qualification testing. Only commercial products and suppliers existed in 1993. For historical reasons, these 'existing products' did not use the most durable material then available. In hindsight, neither had they been adapted nor optimized to combat conditions often experienced in the plants-sagging doors, damaged sealing surfaces, and many thousands of openings and closings per year. Initial attempts to involve the two existing suppliers in efforts to upgrade these seals were unsuccessful. Another suitable supplier had therefore to be found, and a 'new,' COG-owned seal developed; this was completed in 1997. This paper summarizes its testing, along with that of the two existing products. Resistance to aging has been improved significantly. Testing has shown that an accident can be safely withstood after 10 years of service or 40,000 openings-closings, whichever comes first. AECL's Fluid Sealing Technology Unit (FSTU) has invested in the special moulds, test fixtures and other necessary tooling and documentation required to begin commercial manufacture of this new quality product. Accordingly, as with FSTU's other nuclear products such as pump seals, the long-term supply of door seals to CANDU plants is now protected from many external uncertainties-e.g., commercial products being discontinued, materials being changed, companies going out of business. Manufacturing to AECL's detailed specifications is being subcontracted to the new supplier. FSTU is performing the quality surveillance, inspection, testing, and customer service activities concomitant with direct responsibility for supply to the plants. (author)

Materials engineering issues, LMFBR steam generators

International Nuclear Information System (INIS)

Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

1976-01-01

Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

LMFBR safety experiment facility planning and analysis

International Nuclear Information System (INIS)

Stevenson, M.G.; Scott, J.H.

1976-01-01

In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

Methods for U.S. shielding calculations: applications to FFTF and CRBR designs

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Mynatt, F.R.; Disney, R.K.

1978-01-01

The primary components of the U.S. reactor shielding methodology consist of: (1) computer code systems based on discrete ordinates or Monte Carlo radiation transport calculational methods; (2) a data base of neutron and gamma-ray interaction and gamma-ray-production cross sections used as input in the codes; (3) a capability for processing the cross sections into multigroup or point energy formats as required by the codes; (4) large-scale integral shielding experiments designed to test cross-section data or techniques utilized in the calculations; and (5) a ''sensitivity'' analysis capability that can identify the most important interactions in a transport calculation and assign uncertainties to the calculated result that are based on uncertainties in all of the input data. The required accuracy for the methodology is to within 5 to 10% for responses at locations near the core to within a factor of 2 for responses at distant locations. Under these criteria, the methodology has proved to be adequate for in-vessel LMFBR calculations of neutron transport through deep sodium and thick iron and stainless steel shields, of neutron streaming through lower axial coolant channels and primary pipe chaseways, and of the effects of fuel stored within the reactor vessel. For ex-vessel LMFBR problems, the methodology requires considerable improvement, the areas of concern including neutron streaming through heating and ventilation ducts, through the cavity surrounding the reactor vessel, and through gaps around rotating plugs in the reactor heat, as well as gamma-ray streaming through plant shield penetrations

Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

Energy Technology Data Exchange (ETDEWEB)

Himes, D.A.

1997-11-17

The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines.

Swelling of pure copper and copper alloys after high fluence irradiation in FFTF [Fast Flux Test Facility] at approximately 4500C

International Nuclear Information System (INIS)

Garner, F.A.; Brager, H.R.

1986-03-01

The swelling of pure copper and various copper-base alloys has been determined at 47.2 dpa after irradiation in FFTF-MOTA at ∼450 0 C. Data are also becoming available at 63.3 dpa. The alloys tend to fall into two broad categories, those that swell appreciably, sometimes with an S-shaped behavior, and those that resist swelling to very high neutron exposures. It appears that copper may have an intrinsic swelling rate of ∼1%/dpa that is often not reached due to its tendency toward saturation of swelling. The most swelling-resistant alloys examined to date are CuAl25, MZC and Cu-2.0Be

Standard Test Method for Testing Nonmetallic Seal Materials by Immersion in a Simulated Geothermal Test Fluid

CERN Document Server

American Society for Testing and Materials. Philadelphia

1985-01-01

1.1 This test method covers a procedure for a laboratory test for performing an initial evaluation (screening) of nonmetallic seal materials by immersion in a simulated geothermal test fluid. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific precautionary statements, see Section 6 and 11.7.

Hydrodynamic perception in true seals (Phocidae) and eared seals (Otariidae).

Science.gov (United States)

Hanke, Wolf; Wieskotten, Sven; Marshall, Christopher; Dehnhardt, Guido

2013-06-01

Pinnipeds, that is true seals (Phocidae), eared seals (Otariidae), and walruses (Odobenidae), possess highly developed vibrissal systems for mechanoreception. They can use their vibrissae to detect and discriminate objects by direct touch. At least in Phocidae and Otariidae, the vibrissae can also be used to detect and analyse water movements. Here, we review what is known about this ability, known as hydrodynamic perception, in pinnipeds. Hydrodynamic perception in pinnipeds developed convergently to the hydrodynamic perception with the lateral line system in fish and the sensory hairs in crustaceans. So far two species of pinnipeds, the harbour seal (Phoca vitulina) representing the Phocidae and the California sea lion (Zalophus californianus) representing the Otariidae, have been studied for their ability to detect local water movements (dipole stimuli) and to follow hydrodynamic trails, that is the water movements left behind by objects that have passed by at an earlier point in time. Both species are highly sensitive to dipole stimuli and can follow hydrodynamic trails accurately. In the individuals tested, California sea lions were clearly more sensitive to dipole stimuli than harbour seals, and harbour seals showed a superior trail following ability as compared to California sea lions. Harbour seals have also been shown to derive additional information from hydrodynamic trails, such as motion direction, size and shape of the object that caused the trail (California sea lions have not yet been tested). The peculiar undulated shape of the harbour seals' vibrissae appears to play a crucial role in trail following, as it suppresses self-generated noise while the animal is swimming.

LMFBR with booster pump in pumping loop

International Nuclear Information System (INIS)

Rubinstein, H.J.

1975-01-01

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

Seal Related Development Activities at EG/G

Science.gov (United States)

Greiner, Harold F.

1991-01-01

Seal related development activities including modeling, analysis, and performance testing are described for several current seal related projects. Among the current seal related projects are the following: high pressure gas sealing systems for turbomachinery; brush seals for gas path sealing in gas turbines; and tribological material evaluation for wear surfaces in sealing systems.

Status of fast breeder reactor development in the United States

Energy Technology Data Exchange (ETDEWEB)

Horton, K [U.S. Department of Energy, Washington, DC (United States)

1981-05-01

The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

Status of fast breeder reactor development in the United States

International Nuclear Information System (INIS)

Horton, K.

1981-01-01

The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

Demonstration of a dynamic clearance seal in a rotating test facility.

OpenAIRE

Messenger, Andrew; Williams, Richard; Ingram, Grant; Hogg, Simon; Tibos, Stacie; Seaton, Jon; Charnley, Bernard

2017-01-01

The successful demonstration of the “Aerostatic Seal” in a half scale rotating facility is described in this paper. The Aerostatic seal is a novel dynamic clearance seal specifically designed for steam turbine secondary gas path applications. The seal responds to radial rotor excursions, so a reduced clearance can be maintained compared to conventional labyrinth seal without damage to the seal. This enables increased turbine performance through reduced leakage and increased tolerance of turbi...

Specialists meeting on LMFBR flow induced vibrations. Summary report

International Nuclear Information System (INIS)

1977-12-01

A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs

Specialists meeting on LMFBR flow induced vibrations. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-01

A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs.

MIT LMFBR blanket research project. Final summary report

International Nuclear Information System (INIS)

Driscoll, M.J.

1983-08-01

This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

Evaluation of bundle duct interaction by out of pile compressive test of FBR bundles. FFTF type bundle

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2000-10-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle becomes one of the main limiting factors for fuel life times. Then, it is important for the design of fast reactor fuel assembly to understand the BDI behavior in detail. In order to understand the BDI behavior, out of pile compressive tests were conducted for FFTF type bundle by use of X-ray CT equipment. In these compressive tests, two type bundles with different accuracy of initial wire position were conducted. The objective of this test is to evaluate the influence of the initial error from standard position of wire at the same axial position. The locations of the pins and the duct flats are analyzed from CT image data. Quantitative evaluation was performed at the CT image data and discussed the bundle deformation status under BDI condition. Following results are obtained. 1) The accuracy of initial wire position is strongly depends on the pin-to-duct contact behavior. In the case of bundle with large error from standard position, pin-to-duct contact is delayed. 2) The BDI mitigation of the bundle with small error from standard wire position is following: The elastic ovality is the dominant deformation in mild BDI condition, then the wire dispersion and pin dispersion are occurred in severe BDI condition. 3) The BDI mitigation of the bundle with large error from standard wire position is following: The elastic ovality and local bowing of pins with large error from standard wire position are occurred in mild BDI condition, then pin dispersion is occurred around pins with large error from standard wire position, finally wire dispersion is occurred in severe BDI condition. 4) The existence of pins with large error from standard wire position is effective to delay the pin-to-duct contact, but the existence of these pins is possible to contact of pin- to- pin. (author)

The radiological significance of transuranium radioisotopes released to the environment during operation of the LMFBR fuel cycle

International Nuclear Information System (INIS)

Barr, N.F.

1976-01-01

Estimates based on current knowledge and conservative assumptions indicate that release of transuranium elements from the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycle are likely to proaduce population dose commitments small compared to those produced by naturally occurring alpha emitters and globally dispersed transuranium radioisotopes from tests of nuclear weapons in the atmosphere. Potential health consequences of these releases to current and future generations are estimated to be very small compared to risks associated with the production of energy by fossil fuels. The estimates are subject to a number of uncertainties imposed by lack of knowledge. Some of the uncertainties are not likely to be greatly reduced until LMFBR facilities are designed and operated. Others may be significantly reduced prior to facility design and operation. The paper discusses the sensitivity of the estimates to uncertainties and approches to reducing those uncertainties that strongly influence the estimates. (author)

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

International Nuclear Information System (INIS)

Todreas, N.E.; Golay, M.W.; Wolf, L.

1977-01-01

Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

Reactor coolant pump shaft seal behavior during station blackout

International Nuclear Information System (INIS)

Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

1985-04-01

A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

German position paper on structural analysis for LMFBR applications

International Nuclear Information System (INIS)

Angerbauer, A.; Link, F.

1983-01-01

During the design period of the German LMFBR, the SNR-300, extensive work had been done in the field of elastic and inelastic analysis. Furthermore, special design rules have been developed. A review of these activities and their state-of-the art is outlined in this paper

Work plan: transient release from LMFBR fuel

International Nuclear Information System (INIS)

Kress, T.S.; Parker, G.W.; Fontana, M.H.

1975-09-01

The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

International Nuclear Information System (INIS)

Minden, C. v.; Schultheiss, G.F.

1982-01-01

A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

Input parameters to codes which analyze LMFBR wire-wrapped bundles

International Nuclear Information System (INIS)

Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

1980-12-01

This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs

Inferences from new plant design from fast flux test facility operation

International Nuclear Information System (INIS)

Peterson, R.E.; Peckinpaugh, C.L.; Simpson, D.E.

1985-04-01

Experience gained through operation of the Fast Flux Test Facility (FFTF) is now sufficiently extensive that this experience can be utilized in designing the next generation of liquid metal fast reactors. Experience with FFTF core and plant components is cited which can result in design improvements to achieve inherently safe, economic reactor plants. Of particular interest is the mixed oxide fuel system which has demonstrated large design margins. Other plant components have also demonstrated high reliability and offer capital cost reduction opportunities through design simplifications. The FFTF continues to be a valuable US resource which affords prototypic development and demonstration, contributing to public acceptability of future plants

Spray sealing: A breakthrough in integral fuel tank sealing technology

Science.gov (United States)

Richardson, Martin D.; Zadarnowski, J. H.

1989-11-01

In a continuing effort to increase readiness, a new approach to sealing integral fuel tanks is being developed. The technique seals potential leak sources by spraying elastomeric materials inside the tank cavity. Laboratory evaluations project an increase in aircraft supportability and reliability, an improved maintainability, decreasing acquisition and life cycle costs. Increased usable fuel volume and lower weight than conventional bladders improve performance. Concept feasibility was demonstrated on sub-scale aircraft fuel tanks. Materials were selected by testing sprayable elastomers in a fuel tank environment. Chemical stability, mechanical properties, and dynamic durability of the elastomer are being evaluated at the laboratory level and in sub-scale and full scale aircraft component fatigue tests. The self sealing capability of sprayable materials is also under development. Ballistic tests show an improved aircraft survivability, due in part to the elastomer's mechanical properties and its ability to damp vibrations. New application equipment, system removal, and repair methods are being investigated.

LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

International Nuclear Information System (INIS)

Adams, S.R.

1985-10-01

A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences

Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

Energy Technology Data Exchange (ETDEWEB)

Adams, S.R.

1985-10-01

A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

Comparison of the SASSYS/SAS4A radial core expansion reactivity feedback model and the empirical correlation for FFTF

International Nuclear Information System (INIS)

Wigeland, R.A.

1987-01-01

The present emphasis on inherent safety for LMR designs has resulted in a need to represent the various reactivity feedback mechanisms as accurately as possible. The dominant negative reactivity feedback has been found to result from radial expansion of the core for most postulated ATWS events. For this reason, a more detailed model for calculating the reactivity feedback from radial core expansion has been recently developed for use with the SASSYS/SAS4A Code System. The purpose of this summary is to present an extension to the model so that it is more suitable for handling a core restraint design as used in FFTF, and to compare the SASSYS/SAS4A results using this model to the empirical correlation presently being used to account for radial core expansion reactivity feedback to FFTF

Actively controlled shaft seals for aerospace applications

Science.gov (United States)

Salant, Richard F.

1995-07-01

This study experimentally investigates an actively controlled mechanical seal for aerospace applications. The seal of interest is a gas seal, which is considerably more compact than previous actively controlled mechanical seals that were developed for industrial use. In a mechanical seal, the radial convergence of the seal interface has a primary effect on the film thickness. Active control of the film thickness is established by controlling the radial convergence of the seal interface with a piezoelectric actuator. An actively controlled mechanical seal was initially designed and evaluated using a mathematical model. Based on these results, a seal was fabricated and tested under laboratory conditions. The seal was tested with both helium and air, at rotational speeds up to 3770 rad/sec, and at sealed pressures as high as 1.48 x 10(exp 6) Pa. The seal was operated with both manual control and with a closed-loop control system that used either the leakage rate or face temperature as the feedback. The output of the controller was the voltage applied to the piezoelectric actuator. The seal operated successfully for both short term tests (less than one hour) and for longer term tests (four hours) with a closed-loop control system. The leakage rates were typically 5-15 slm (standard liters per minute), and the face temperatures were generally maintained below 100C. When leakage rate was used as the feedback signal, the setpoint leakage rate was typically maintained within 1 slm. However, larger deviations occurred during sudden changes in sealed pressure. When face temperature was used as the feedback signal, the setpoint face temperature was generally maintained within 3 C, with larger deviations occurring when the sealed pressure changes suddenly. the experimental results were compared to the predictions from the mathematical model. The model was successful in predicting the trends in leakage rate that occurred as the balance ratio and sealed pressure changed

SEALING SIMULATED LEAKS

Energy Technology Data Exchange (ETDEWEB)

Michael A. Romano

2004-09-01

This report details the testing equipment, procedures and results performed under Task 7.2 Sealing Simulated Leaks. In terms of our ability to seal leaks identified in the technical topical report, Analysis of Current Field Data, we were 100% successful. In regards to maintaining seal integrity after pigging operations we achieved varying degrees of success. Internal Corrosion defects proved to be the most resistant to the effects of pigging while External Corrosion proved to be the least resistant. Overall, with limitations, pressure activated sealant technology would be a viable option under the right circumstances.

Assessment of accident energetics in LMFBR core-disruptive accidents

International Nuclear Information System (INIS)

Fauske, H.K.

1977-01-01

An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

International Nuclear Information System (INIS)

Roake, W.E.

1977-01-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

1977-04-01

Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

Shielding plug for LMFBR type reactors

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1979-01-01

Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

Cask systems development program seal technology

International Nuclear Information System (INIS)

Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

1991-01-01

General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (10 CFR 71). Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. Experiments were performed to characterize the performance of several seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Results show that the seal materials tested, with the exception of silicone S613-60, are not leak tight (1 x 10 -7 std cm 3 /s) at manufacturer low-temperature ratings

Design modification of the in-pile test section for increase of sealing capability

Energy Technology Data Exchange (ETDEWEB)

Hong, J T; Ahn, S H; Joung, C Y; Jeong, H Y; Lee, J M; Sim, B S [Department of Research Reactor Utilization and Development, Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2012-03-15

Since KAERI established the fuel test loop (FTL) at HANARO in 2009, KAERI has carried out several experiments to verify the performances of the equipment. Based on the experiments, the design modification of the In-Pile test Section (IPS) has been processed to improve some difficulties such as difficulty in ejecting the inner assembly of the IPS from the pressure vessel, difficulty of the sealing process of the cooling water, etc. At first, because the cooling water of HANARO in KAERI consists of an open-pool type, if a certain shock is generated during the disassembly process, the cooling water can be spattered out of the pool. Therefore, two jacking bolts will be added on the top flange part of the inner assembly to decrease the shock. Second, at the pressure boundary of the IPS where MI-cables go through, the brazing process has been used to seal out the cooling water. However, because the length of the IPS is up to 5.5 meters, it is too difficult and time consuming to carry out the brazing process at the end part of the IPS. Therefore, the brazing process will be replaced with the mechanical sealing structure to simplify the assembly process. (author)

Structural analysis for LMFBR applications[Indian position paper

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-05-01

Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start.

Effect of operating temperature on LMFBR core performance

International Nuclear Information System (INIS)

Noyes, R.C.; Bergeron, R.J.; di Lauro, G.F.; Kulwich, M.R.; Stuteville, D.W.

1977-01-01

The purpose of the study is to provide an engineering evaluation of high and low temperature LMFBR core designs. The study was conducted by C-E supported by HEDL expertise in the areas of materials behavior, fuel performance and fabrication/fuel cycle cost. The evaluation is based primarily on designs and analyses prepared by AI, GE and WARD during Phase I of the PLBR studies

Leak detection of KNI seals

International Nuclear Information System (INIS)

Baranyai, G.; Peter, A.; Windberg, P.

1990-03-01

In Unit 3 and 4 of the Paks Nuclear Power Plant, Hungary, KNI type seals are used as lead-throughs with conical nickel sealing rings. Their failure can be critical for the operation of the reactor. An Acoustical Leak Detection System (ALDS) was constructed and tested for the operational testing of the seals. Some individual papers are presented in this collection on the calibration and testing of the ALDS intended to be placed on the top of the reactor vessels. The papers include simulation measurements of Unit 3 of NPP, laboratory experiments, evaluation of measurements, and further development needs with the ALDS. (R.P.) 50 figs.; 19 tabs

Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

International Nuclear Information System (INIS)

Wootan, D.W.; Omberg, R.P.

2015-01-01

Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

1974-01-01

The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

Confirmatory simulation of safety and operational transients in LMFBR systems

International Nuclear Information System (INIS)

Guppy, J.G.; Agrawal, A.K.

1978-01-01

Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated

Performance testing of prototype live loaded packed stem seals for large gate valves in pressurized hot water

International Nuclear Information System (INIS)

Pothier, N.E.

1976-01-01

Prototype live loaded packed stem seals for large gate valves have been tested in a laboratory. The test fluid was demineralized water at 547 K, 8.27 MPa and pH 10. Nine packing configurations were tested; three different commercial brands of asbestos/graphite valve packings and three different sizes for each packing brand. Conventional and live loaded packed stem seals are briefly described. Stem leakage, packing consolidation and stem friction data are given. For all tests, leakage rates of less than 10 g d -1 were observed. It was also observed that stem friction was significantly affected by thermal expansion of the stem. (author)

Fast Flux Test Facility: first three years of operation and future mission

International Nuclear Information System (INIS)

Peckinpaugh, C.L.; Newland, D.J.; Evans, E.A.

1985-03-01

In summary, the FFTF has proven to be a high performance, versatile test reactor. Results obtained during its first three years of operation - and those to be obtained in the coming years - are building a technology and experience base that is invaluable to future LMRs. The FFTF demonstrates proven LMR technology with a focus for the future and provides the US with international LMR technology leadership

Ferritic steels for French LMFBR steam generators

International Nuclear Information System (INIS)

Aubert, M.; Mathieu, B.; Petrequin, P.

1983-06-01

Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

Tests of Shaft Seal Systems of Circulation Pumps during Station Blackout

Energy Technology Data Exchange (ETDEWEB)

Beisiegel, A.; Foppe, F.; Wich, M.

2014-07-01

AREVA GmbH operates a unique Thermal-hydraulic plat form in Germany, France and USA. It is recognised as a test body according to ISO 17025. The Deutsche Akkreditierungsstelle GmbH (DAkkS - German Society for Accreditation) has also certified the Thermal-hydraulic platform as an independent inspection body Type C according to ISO 17020. A part of this platform is the Component Laboratory located in Karlstein, Germany which is in operation since more than 50 years. The testing activities cover a wide range as: Critical Heat Flux Tests, Valve Testing and Environmental Qualification for safety related components. Since 2011 the Component Qualification Karlstein extended their testing scope for different types of Shaft Seal Systems. (Author)

Investigating the sealing capacity of a seal system in rock salt (DOPAS project)

Energy Technology Data Exchange (ETDEWEB)

Jantschik, Kyra; Moog, Helge C.; Czaikowski, Oliver; Wieczorek, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Braunschweig (Germany)

2016-11-15

This paper describes research and development work on plugging and sealing repositories, an issue of fundamental importance for the rock salt option which represents one of the three European repository options, besides the clay rock and the crystalline rock options. The programme aims at providing experimental data needed for the theoretical analysis of the long-term sealing capacity of concrete- based sealing materials. In order to demonstrate hydro-mechanical material stability under representative load scenarios, a comprehensive laboratory testing programme is carried out. This comprises investigation of the sealing capacity of the combined seal system and impact of the so-called excavation-damaged zones (EDZ) as well as investigation of the hydro-chemical long-term stability of the seal in contact with different brines under diffusive and advective conditions. This paper presents experimental approaches and preliminary results from laboratory investigations on salt concrete and combined systems as obtained to date.

Testing of improved polyimide actuator rod seals at high temperature and under vacuum conditions for use in advanced aircraft hydraulic systems

Science.gov (United States)

Sellereite, B. K.; Waterman, A. W.; Nelson, W. G.

1974-01-01

Polyimide second-stage rod seals were evaluated to determine their suitability for applications in space station environments. The 6.35-cm (2.5-in.)K-section seal was verified for thermal cycling operation between room temperature and 478 K (400 F) and for operation in a 133 micron PA(0.000001 mm Hg) vacuum environment. The test seal completed the scheduled 96 thermal cycles and 1438 hr in vacuum with external rod seal leakage well within the maximum allowable of two drops per 25 actuation cycles. At program completion, the seals showed no signs of structural degradation. Posttest inspection showed the seals retained a snug fit against the shaft and housing walls, indicating additional wear life capability. Evaluation of a molecular flow section during vacuum testing, to inhibit fluid loss through vaporization, showed it to be beneficial with MIL-H-5606, a petroleum-base fluid, in comparison with MIL-H-83282, a synthetic hydrocarbon-base fluid.

Recent advances in magnetic liquid sealing

International Nuclear Information System (INIS)

Raj, K.; Stahl, P.; Bottenberg, W.; True, D.; Martis, G.; Zook, C.

1979-01-01

In this paper recent work in design and testing of two special magnetic liquid seals extending the state-of-the-art of ferrofluidic sealing is discussed. These custom seals are a moving belt edge seal and an exclusion seal. The first seal provides a hermetic barrier to solid particulates expected to be present in enclosed nuclear environments. The second seal is used on a magnetic disk drive spindle and reduces the particulate contaminants in the memory disk pack area by up to three orders of magnitude. In addition, bearing life in the spindle is found to be doubled due to reduction of operating temperature. The fundamentals of magnetic fluid sealing are presented in terms of magnetic circuit design and physical properties of ferrofluids

A critical experimental study of integral physics parameters in simulated LMFBR meltdown cores

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Wade, D.C.; Bucher, R.G.; Smith, D.M.; McKnight, R.D.; Lesage, L.G.

1978-01-01

Integral physics parameters of several representative, idealized meltdown LMFBR configurations were measured in mockup critical assemblies on the ZPR-9 reactor at Argonne National Laboratory. The experiments were designed to provide data for the validation of analytical methods used in the neutronics part of LMFBR accident analysis. Large core distortions were introduced in these experiments (involving 18.5% core volume) and the reactivity worths of configuration changes were determined. The neutronics parameters measured in the various configurations showed large changes upon core distortion. Both diffusion theory and transport theory methods were shown to mispredict the experimental configuration eigenvalues. In addition, diffusion theory methods were shown to result in a non-conservative misprediction of the experimental configuration change worths. (author)

Design approaches to achieve competitive LMFBR capital costs

International Nuclear Information System (INIS)

Arnold, W.H.; Ehrman, C.S.; Sharbaugh, J.E.; Young, W.H.

1982-01-01

Through analysis of the essential functional elements of an LMFBR, numerous ways were found to simplify system design, reduce the size of components and equipment, and eliminate some components and systems. The projected capital cost per net kW of this design is competitive with that of current PWRs. RandD programs and the construction and operation of CRBRP now are needed to prove out the features of this new design

Compliant seal development

Science.gov (United States)

Hendricks, Robert C.

1993-10-01

The compliant metallic seal combines the noncontact feature of the labyrinth seal, the low leakage of a mechanical seal, and the compliant nature of the brush seal. It consists of several thin metallic elements or leaves mounted within a ring which is press fit into the housing, and in form, sort of resembles a lip seal sections wiping the shaft. A second set of overlapping cover leaves are placed on top of the shaft riding leaves which reduces leakage and provides stiffness. The leaves can be straight or angle cut. The shaft riding fingers are designed with mismatched curvature to provide lift off similar to the Rayleigh lift pads in mechanical seals with leading edge clearances nearly twice those of the trailing edge as as shown by Fleming to be optimal for gas flows in convergent seal passages. Leading edge clearances range from 300 to 500 microinches. Balance pockets beneath the leaves provide fluid film feed to the 'Rayleigh lift' surface and the proper balance ratio (mechanical seal) when combined with the static pressure and film pressure. The leaves flex in the radial direction and accommodate thermomechanical behavior as well as axial motion and angular misalignment. In the static mode, there is a net closing force on the leaves. The seals were tested to 70 psi at speeds to 16,000 rpm or surface speeds to 330 fps and temperatures from ambient to 440 F. A slow cycle through the rig critical at 10,000 rpm induced a radial vibration response of 0.004 to 0.005 inch were accommodated by the seal. Preliminary performance data are encouraging demonstrating hydrodynamic liftoff and noncontacting operation at pressure and speeds typical of gas turbine engines. The leakage performance data are significantly better than commercial labyrinth and brush seals which should be expected as this design incorporates the features of the low leakage face or mechanical seal along with the flexibility of the brush configuration.

The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2007-10-15

The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

Study of structural attachments of a pool type LMFBR vessel through seismic analysis of a simplified three dimensional finite element model

International Nuclear Information System (INIS)

Ahmed, H.; Ma, D.

1979-01-01

A simplified three dimensional finite element model of a pool type LMFBR in conjunction with the computer program ANSYS is developed and scoping results of seismic analysis are produced. Through this study various structural attachments of a pool type LMFBR like the reactor vessel skirt support, the pump support and reactor shell-support structure interfaces are studied. This study also provides some useful results on equivalent viscous damping approach and some improvements to the treatment of equivalent viscous damping are recommended. This study also sets forth pertinent guidelines for detailed three dimensional finite element seismic analysis of pool type LMFBR

Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

International Nuclear Information System (INIS)

2002-01-01

The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

Test results on the dynamic testing of expansion type concrete anchors

International Nuclear Information System (INIS)

Barron, B.; Rice, R.; Stephen, R.M.

1974-12-01

Tests were performed to determine the structural response of commercially available expansion type anchors for the FFTF when subjected to dynamic loadings similar to machine vibrations and earthquakes. The specimens were subjected to tension, shear, and combined shear-tension loads

Inflatable Module Seal Interface Development and Testing

Data.gov (United States)

National Aeronautics and Space Administration — Develop a repeatable low permeable sealing interface evaluating O-ring, RTV bond and flowed RTV bond methods. Advanced Bladder materials (ArmorFlex, Nanoclay, etc)...

Turbine Seal Research at NASA GRC

Science.gov (United States)

Proctor, Margaret P.; Steinetz, Bruce M.; Delgado, Irebert R.; Hendricks, Robert C.

2011-01-01

Low-leakage, long-life turbomachinery seals are important to both Space and Aeronautics Missions. (1) Increased payload capability (2) Decreased specific fuel consumption and emissions (3) Decreased direct operating costs. NASA GRC has a history of significant accomplishments and collaboration with industry and academia in seals research. NASA's unique, state-of-the-art High Temperature, High Speed Turbine Seal Test Facility is an asset to the U.S. Engine / Seal Community. Current focus is on developing experimentally validated compliant, non-contacting, high temperature seal designs, analysis, and design methodologies to enable commercialization.

Design and Calibration of a Full Scale Active Magnetic Bearing Based Test Facility for Investigating Rotordynamic Properties of Turbomachinery Seals in Multiphase Flow

DEFF Research Database (Denmark)

Voigt, Andreas Jauernik; Mandrup-Poulsen, Christian; Nielsen, Kenny Krogh

2016-01-01

of the subsea installed rotating machinery. It is well known that careful design of turbomachinery seals, such as interstage and balance piston seals, is pivotal for the performance of pumps and compressors. Consequently, the ability to predict the complex interaction between fluid dynamics and rotordynamics...... University of Denmark and Lloyd's Register Consulting are currently establishing a purpose built state of the art multiphase seal test facility, which is divided into three modules. Module I consists of a full scale Active Magnetic Bearing (AMB) based rotor dynamic test bench. The internally designed custom...... AMBs are equipped with an embedded Hall sensor system enabling high precision non-contact seal force quantification. Module II is a fully automatised calibration facility for the Hall sensor based force quantification system. Module III consists of the test seal housing assembly. This paper provides...

Electronically Controlled Mechanical Seal for Aerospace Applications -- Part 1: Design, Analysis, and Steady State Tests

Science.gov (United States)

Salant, Richard F.; Wolff, Paul; Navon, Samuel

1994-01-01

An electronically-controlled mechanial seal, for use as the purge gas seal in a liquid oxygen turbopump, has been designed, analyzed, and built. The thickness of the lubricating film between the faces is controlled by adjusting the coning of the carbon face. This is done by applying a voltage across a piezoelectric element to which the carbon face is bound. Steady state tests have shown that the leakage rate (and film thickness) can be adjusted over a substantial range, utilizing the available range of voltage.

Electronically controled mechanical seal for aerospace applications -- Part 1: Design, analysis, and steady state tests

Science.gov (United States)

Salant, Richard F.; Wolff, Paul; Navon, Samuel

1994-01-01

An electronically-controlled mechanial seal, for use as the purge gas seal in a liquid oxygen turbopump, has been designed, analyzed, and built. The thickness of the lubricating film between the faces is controlled by adjusting the coning of the carbon face. This is done by applying a voltage across a piezoelectric element to which the carbon face is bound. Steady state tests have shown that the leakage rate (and film thickness) can be adjusted over a substantial range, utilizing the available range of voltage.

Turbine and Structural Seals Team Facilities

Data.gov (United States)

Federal Laboratory Consortium — Seals Team Facilities conceive, develop, and test advanced turbine seal concepts to increase efficiency and durability of turbine engines. Current projects include...