[en] A loss-of-coolant test sponsored jointly by Ontario Hydro and Atomic Energy of Canada Limited (AECL) will be performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The test will provide both an in-reactor evaluation of CANDU fuel element thermal mechanical behavior during a postulated Loss-of-Coolant Accident and an experimental data base for evaluation of the Canadian transient fuel behavior code, ELOCA. The PFB-CANDU test train is comprised of four independently shrouded CANDU design fuel elements installed in symmetric orientations in a modified PBF test assembly. Three of the fuel elements were preirradiated in the NRX reactor to approximately 5000 MWd/t (120 MWh/kgU) at approx. 53 kW/m, and the fourth fuel element is unirradiated

[en] A loss-of-coolant test sponsored jointly by Ontario Hydro and Atomic Energy of Canada Limited (AECL) will be performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The test will provide both an in-reactor evaluation of CANDU fuel element thermal mechanical behavior during a postulated Loss-of-Coolant Accident and an experimental data base for evaluation of the Canadian transient fuel behavior code, ELOCA. The PBF-CANDU test train is comprised of four independently shrouded CANDU design fuel elements installed in symmetric orientations in a modified PBF test assembly. Three of the fuel elements were preirradiated in the NRX reactor to approximately 5000 MWd/t (120 MWh/kgU) at approx. 53 kW/m, and the fourth fuel element is unirradiated. RELAP5 was used to simulate the PBF-CANDU Loss-of-Coolant test. The RELAP5 calculations were used to determine the blowdown valve sequencing required to obtain the prescribed depressurization. The FRAP-T6 code was used to calculate the fuel element response to the simultaneous blowdown and power excursion during the test

[en] There has been very little, documented decontamination and decommissioning (D ampersand D) experience on which to project cleanup costs and schedules for plutonium facilities at SRS and other DOE sites. A portion of the HB-Line, a plutonium-238 processing facility at SRS, has been undergoing D ampersand D intermittently since 1984. Although this cleanup effort was not originally intended to quantify results, some key data have been project has demonstrated effective methods of accumulated, and the performing D ampersand D work, and has demonstrated cleanup equipment and techniques under conditions of high contamination. Plutonium facilities where D ampersand D is already underway provide an opportunity for' timely field testing of characterization, size reduction, and decontamination techniques. Some data are presented here; however, more specific tests and data may be obtained during the remainder of this project. This project has been recommended as a candidate test facility for a DOE planned ''Integrated D ampersand D Demonstration'' managed by EM-50 to develop and demonstrate technology for D ampersand D and surplus facilities deactivation. Both the remainder of this project and the Integrated D ampersand D Demonstration Program can benefit from a joint effort, and the, overall costs should be reduced

No abstract available

[en] The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure

[en] The Severe Core Damage Analysis Package (SCDAP) computer code is being developed at the Idaho National Engineering Laboratory under the sponsorship of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. SCDAP models the progression of light water reactor core damage including core heatup, core disruption, debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data

[en] The Thermal Fuels Behavior Program of EG and G Idaho, Inc., has conducted irradiation tests since 1975 for the U.S. Nuclear Regulatoy Commission to determine the effects of operation in overpower or undercooling modes on pressurized water reactor (PWR) type fuel rods. Highly instrumented single-rod tests, performed in the power-cooling-mismatch (PCM) test series, have produced extensive data on the coolant thermal-hydraulic conditions required to produce departure from nuclear boiling (DNB) and the consequences to Zircaloy cladding integrity of sustained film boiling operation. Test PCM-5, the first nine-rod cluster test of the PCM test series, (a 3 x 3 rod array within a flow shroud) was performed in May 1978 to (a) assess film boiling behavior of a fuel rod in a configuration representative of power reactor fuel bundles, and (b) to determine whether rod-to-rod DNB propagation or failure propagation would occur. Test results are discussed

[en] The three contract types used by US DOE customers are described: requirement, long-term fixed commitment, and adjustable fixed commitment. The early 1979 enrichment planning estimate of domestic and non-US nuclear power on-line in 1985 which would obtain enrichment services from DOE is reviewed. Effects of the uncertainty in the nuclear market are discussed. Nomographs are presented for finding the normal uranium feed requirement and separative work requirement

No abstract available

[en] The primary objectives of Test RIA 1-2 were to (a) characterize the response of preirradiated fuel rods during an RIA event conducted at boiling water reactor (BWR) hot-startup conditions for an axial peak pellet surface energy of 200 cal/g UO₂, and (b) evaluate the effect of internal rod pressure on preirradiated fuel rod response during an RIA event. The test consisted of four, individually shrouded, pressurized water reactor-type fuel rods previously irradiated to burnups of about 4800 MWd/t. In addition to the power calibration and preconditioning, the fuel rods were subjected to a single power burst that deposited a total pellet surface energy of approximately 200 cal/gm UO₂ at the axial peak power location (estimated using the core power chambers to relate steady state and transient powers). The test data indicate that the two irradiated fuel rods prepressurized to 2.41 MPa did not fail. FRAP-T4 calculations had predicted that prompt cladding rupture would occur for pellet surface energy depositions of 206 cal/g or greater. Although the two fuel rods prepressurized to 2.41 MPa did not fail, the data indicate that at least one of the two fuel rods prepressurized to 0.1 MPa did fail. Based on the core power chamber data, this rod failure indicates a threshold for the preirradiated fuel rods near or below 200 cal/g UO₂ total pellet surface energy at the axial flux peak