The Experts below are selected from a list of 48 Experts worldwide ranked by ideXlab platform

M.r. Lindquist - One of the best experts on this subject based on the ideXlab platform.

  • summary of an assessment of potential waterborne accidents during transport of the Shippingport Reactor pressure vessel and neutron shield tank assembly
    Nuclear Materials Management. Annual Meeting Proceedings; (United States), 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, was constructed in the mid-1950's to demonstrate pressurized water Reactor technology and to generate electric power. The Shippingport Reactor reached the end of its design life in 1984 and was turned over to the US Department of Energy of decommissioning. The Reactor pressure vessel/neutron shield tank was removed as a unit and transported by barge from the Shippingport Project Site on the Ohio River to the US Department of Energy Hanford Site in Richland, Washington, for interment. This paper summarizes the work performed to provide supplemental information to the safety analysis report for packaging on hazards associated with transporting the package by water. Potential waterborne accident scenarios were identified along with the probable accident recovery mitigation measures, which could have been initiated, if required. The risk evaluations are based on an assessment of the probabilities of an accident in which the barge does not sink; an accident in which the barge sinks, but with no damage to the Reactor pressure vessel/neutron shield tank; an accident in which the barge sinks, but with possible damage to the package; and for accidents resulting in a postulated fire. 5 refs., 2 figs., 3 tabs.

  • Drop testing conducted to benchmark the Shippingport Reactor pressure vessel package safety analysis
    Nuclear Engineering and Design, 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    Abstract The decommissioned Shippingport Reactor pressure vessel and its integral neutron shield tank were transported from Shippingport, Pennsylvania, via barge to Richland, Washington, for burial in the Hanford Site radioactive waste disposal area. To ensure that the Reactor pressure vessel/neutron shield tank assembly could be shipped safely without undue risk to the public or the environment, the Reactor pressure vessel/neutron shield tank assembly was certified by the U.S. Department of Energy as a type B package. A safety analysis report for packaging was prepared in accordance with U.S. Department of Energy requirements to provide the technical basis for the U.S. Department of Energy certification. The Reactor pressure vessel/neutron shield tank package is a monolithic structure of lightweight concrete and steel. Its estimated weight is 844 t (930 tons). To substantiate multidimensional inelastic analyses, a series of 11 drop tests was conducted on 7 benchmark models from heights of 30.5 cm (1 ft), 9.14 m (30 ft), and 13.7 m (45 ft). Technical evaluation and correlation of the test data were performed in conjunction with the structural analysis and assessment of the package. This paper provides a comprehensive discussion on the benchmark drop models and specific drop tests and also addresses the results obtained from comparing technical data with analytical data.

D.l. Becker - One of the best experts on this subject based on the ideXlab platform.

  • summary of an assessment of potential waterborne accidents during transport of the Shippingport Reactor pressure vessel and neutron shield tank assembly
    Nuclear Materials Management. Annual Meeting Proceedings; (United States), 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, was constructed in the mid-1950's to demonstrate pressurized water Reactor technology and to generate electric power. The Shippingport Reactor reached the end of its design life in 1984 and was turned over to the US Department of Energy of decommissioning. The Reactor pressure vessel/neutron shield tank was removed as a unit and transported by barge from the Shippingport Project Site on the Ohio River to the US Department of Energy Hanford Site in Richland, Washington, for interment. This paper summarizes the work performed to provide supplemental information to the safety analysis report for packaging on hazards associated with transporting the package by water. Potential waterborne accident scenarios were identified along with the probable accident recovery mitigation measures, which could have been initiated, if required. The risk evaluations are based on an assessment of the probabilities of an accident in which the barge does not sink; an accident in which the barge sinks, but with no damage to the Reactor pressure vessel/neutron shield tank; an accident in which the barge sinks, but with possible damage to the package; and for accidents resulting in a postulated fire. 5 refs., 2 figs., 3 tabs.

  • Drop testing conducted to benchmark the Shippingport Reactor pressure vessel package safety analysis
    Nuclear Engineering and Design, 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    Abstract The decommissioned Shippingport Reactor pressure vessel and its integral neutron shield tank were transported from Shippingport, Pennsylvania, via barge to Richland, Washington, for burial in the Hanford Site radioactive waste disposal area. To ensure that the Reactor pressure vessel/neutron shield tank assembly could be shipped safely without undue risk to the public or the environment, the Reactor pressure vessel/neutron shield tank assembly was certified by the U.S. Department of Energy as a type B package. A safety analysis report for packaging was prepared in accordance with U.S. Department of Energy requirements to provide the technical basis for the U.S. Department of Energy certification. The Reactor pressure vessel/neutron shield tank package is a monolithic structure of lightweight concrete and steel. Its estimated weight is 844 t (930 tons). To substantiate multidimensional inelastic analyses, a series of 11 drop tests was conducted on 7 benchmark models from heights of 30.5 cm (1 ft), 9.14 m (30 ft), and 13.7 m (45 ft). Technical evaluation and correlation of the test data were performed in conjunction with the structural analysis and assessment of the package. This paper provides a comprehensive discussion on the benchmark drop models and specific drop tests and also addresses the results obtained from comparing technical data with analytical data.

O K Chopra - One of the best experts on this subject based on the ideXlab platform.

  • mechanical properties of thermally aged cast stainless steels from Shippingport Reactor components
    2. international conference on heat resistant materials Gatlinburg TN (United States) 11-14 Sep 1995, 1995
    Co-Authors: O K Chopra
    Abstract:

    A procedure and correlations are presented for predicting mechanical properties of cast stainless steels in service at temperatures <450{degrees}C from known material information. The ``saturation`` fracture properties of a cast stainless steel, i.e., the minimum values that would be achieved for the material after long-term service, are estimated from the chemical composition of the steel. Fracture properties as a function of time and temperature of service are estimated from the kinetics of embrittlement, which are also determined from chemical composition. The correlations successfully predict fracture toughness, Charpy-impact, and tensile properties of cast stainless steels from the Shippingport-, Ringhals-, and Gundremmingen-Reactor components.

  • radiation embrittlement of the neutron shield tank from the Shippingport Reactor
    1991
    Co-Authors: O K Chopra, W J Shack, S T Rosinski
    Abstract:

    The irradiation embrittlement of neutron shield tank (NST) material (A212 Grade B steel) from the Shippingport Reactor has been characterized. Irradiation increases the Charpy transition temperature (CTT) by 23--28{degrees}C (41--50{degrees}F) and decreases the upper-shelf energy. The shift in CTT is not as severe as that observed in high-flux isotope Reactor (HFIR) surveillance specimens. However, the actual value of the CTT is higher than that for the HFIR data. The increase in yield stress is 51 MPa (7.4 ksi), which is comparable to HFIR data. The NST material is weaker in the transverse orientation than in the longitudinal orientation. Some effects of position across the thickness of the wall are also observed; the CTT shift is slightly greater for specimens from the inner region of the wall. Annealing studies indicate complete recovery from embrittlement after 1 h at 400{degrees}C (752{degrees}F). Although the weld metal is significantly tougher than the base metal, the shifts in CTT are comparable. The shifts in CTT for the Shippingport NST are consistent with the test and Army Reactor data for irradiations at <232{degrees}C (<450{degrees}F) and show very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor (ORR). The effects of irradiation temperature, fluence rate, and neutron flux spectrum are discussed. The results indicate that fluence rate has no effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center dot}s and at the low operating temperatures of the Shippingport NST, i.e., 55{degrees}C (130{degrees}F). This suggests that the accelerated embrittlement of HFIR surveillance samples is most likely due to the relatively higher proportion of thermal neutrons in the HFIR spectrum compared to that for the test Reactors. 28 refs., 25 figs.

D.m. Burgess - One of the best experts on this subject based on the ideXlab platform.

  • summary of an assessment of potential waterborne accidents during transport of the Shippingport Reactor pressure vessel and neutron shield tank assembly
    Nuclear Materials Management. Annual Meeting Proceedings; (United States), 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, was constructed in the mid-1950's to demonstrate pressurized water Reactor technology and to generate electric power. The Shippingport Reactor reached the end of its design life in 1984 and was turned over to the US Department of Energy of decommissioning. The Reactor pressure vessel/neutron shield tank was removed as a unit and transported by barge from the Shippingport Project Site on the Ohio River to the US Department of Energy Hanford Site in Richland, Washington, for interment. This paper summarizes the work performed to provide supplemental information to the safety analysis report for packaging on hazards associated with transporting the package by water. Potential waterborne accident scenarios were identified along with the probable accident recovery mitigation measures, which could have been initiated, if required. The risk evaluations are based on an assessment of the probabilities of an accident in which the barge does not sink; an accident in which the barge sinks, but with no damage to the Reactor pressure vessel/neutron shield tank; an accident in which the barge sinks, but with possible damage to the package; and for accidents resulting in a postulated fire. 5 refs., 2 figs., 3 tabs.

  • Drop testing conducted to benchmark the Shippingport Reactor pressure vessel package safety analysis
    Nuclear Engineering and Design, 1991
    Co-Authors: D.l. Becker, D.m. Burgess, M.r. Lindquist
    Abstract:

    Abstract The decommissioned Shippingport Reactor pressure vessel and its integral neutron shield tank were transported from Shippingport, Pennsylvania, via barge to Richland, Washington, for burial in the Hanford Site radioactive waste disposal area. To ensure that the Reactor pressure vessel/neutron shield tank assembly could be shipped safely without undue risk to the public or the environment, the Reactor pressure vessel/neutron shield tank assembly was certified by the U.S. Department of Energy as a type B package. A safety analysis report for packaging was prepared in accordance with U.S. Department of Energy requirements to provide the technical basis for the U.S. Department of Energy certification. The Reactor pressure vessel/neutron shield tank package is a monolithic structure of lightweight concrete and steel. Its estimated weight is 844 t (930 tons). To substantiate multidimensional inelastic analyses, a series of 11 drop tests was conducted on 7 benchmark models from heights of 30.5 cm (1 ft), 9.14 m (30 ft), and 13.7 m (45 ft). Technical evaluation and correlation of the test data were performed in conjunction with the structural analysis and assessment of the package. This paper provides a comprehensive discussion on the benchmark drop models and specific drop tests and also addresses the results obtained from comparing technical data with analytical data.

Chopra O. K. - One of the best experts on this subject based on the ideXlab platform.

  • Mechanical properties of thermally aged cast stainless steels from Shippingport Reactor components.
    Argonne National Laboratory, 1995
    Co-Authors: Chopra O. K., Shack W. J., Technology Energy
    Abstract:

    Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport Reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approx}13 y at {approx}281 C (538 F) for the hot-leg components and {approx}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and JIC of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 Reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y and the KRB Reactor pump cover plate (CF-8) after {approx}8 y of service

  • Mechanical properties of thermally aged cast stainless steels from Shippingport Reactor components
    Argonne National Laboratory, 1995
    Co-Authors: Chopra O. K., Shack W. J.
    Abstract:

    Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport Reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approximately}13 y at {approximately}281 C (538 F) for the hot-leg components and {approximately}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 Reactor hot and crossover-leg elbows (CF-8M steel) after service of {approximately} 15 y and the KRB Reactor pump cover plate (CF-8) after {approximately} 8 y of service

  • Studies of aged cast stainless steel from the Shippingport Reactor
    Argonne National Laboratory. Materials and Components Technology Division., 1991
    Co-Authors: Chopra O. K.
    Abstract:

    The mechanical properties of cast stainless steels from the Shippingport Reactor have been characterized. Baseline properties for unaged materials were obtained from tests on either recovery-annealed material or material from a cooler region of the component. The materials exhibited modest decrease in impact energy and fracture toughness and a small increase in tensile strength. The fracture toughness J-R curve, J{sub IC} value, tensile flow stress, and Charpy-impact energy of the materials showed very good agreement with estimations based on accelerated laboratory aging studies. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy that would be achieved after long-term aging, were established from materials that were aged further in the laboratory at temperatures between 320 and 400{degree}C. The results showed very good agreement with estimates; the activation energies ranged from 125 to 250 kJ/mole and the minimum room-temperature impact energy was >75 J/cm{sup 2}. The estimated impact energy and fracture toughness J-R curve for materials from the Ringhals Reactor hot and crossover-leg elbows are also presented