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End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

Description: Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.
Date: October 1, 1987
Creator: Richardson, K.D.
Partner: UNT Libraries Government Documents Department

History of the production complex: The methods of site selection

Description: Experience taught the Atomic Energy Commission how to select the best possible sites for its production facilities. AEC officials learned from the precedents set by the wartime Manhattan Project and from their own mistakes in the immediate postwar years. This volume discusses several site selections. The sites covered are: (1) the Hanford Reservation, (2) the Idaho reactor site, (3) the Savannah River Plant, (4) the Paducah Gaseous Diffusion Plant, (5) the Portsmouth Gaseous Diffusion Plant, (6) the Fernald Production Center, (7) the PANTEX and Spoon River Plants, (8) the Rocky Flats Fabrication Facility, and (9) the Miamisburg and Pinellas plants. (JDH)
Date: September 1, 1987
Partner: UNT Libraries Government Documents Department

A comparison of propulsion systems for potential space mission applications

Description: A derivative of the NERVA nuclear rocket engine was compared with a chemical propulsion system and a nuclear electric propulsion system to assess the relative capabilities of the different propulsion system options for three potential space missions. The missions considered were (1) orbital transfer from low earth orbit (LEO) to geosynchronous earth orbit (GEO), (2) LEO to a lunar base, and (3) LEO to Mars. The results of this comparison indicate that the direct-thrust NERVA-derivative nuclear rocket engine has the best performance characteristics for the missions considered. The combined high thrust and high specific impulse achievable with a direct-thrust nuclear stage permits short operating times (transfer times) comparable to chemical propulsion systems, but with considerably less required propellant. While nuclear-electric propulsion systems are more fuel efficient than either direct-nuclear or chemical propulsion, they are not stand-alone systems, since their relatively low thrust levels require the use of high-thrust ferry or lander stages in high gravity applications such as surface-to-orbit propulsion. The extremely long transfer times and inefficient trajectories associated with electric propulsion systems were also found to be a significant drawback.
Date: January 1, 1987
Creator: Harvego, E.A. & Sulmeisters, T.K.
Partner: UNT Libraries Government Documents Department

Detection and characterization of flaws in segments of light water reactor pressure vessels

Description: Studies have been conducted to determine flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessil and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication (with a through-wall dimension of approx.6 mm (approx.0.24 in.)) was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications (i.e., for a total of approximately 6.8 m/sup 2/ (72 ft/sup 2/) of cladding surface).
Date: January 1, 1987
Creator: Cook, K.V.; Cunningham, R.A. Jr. & McClung, R.W.
Partner: UNT Libraries Government Documents Department

Nuclear reactor power for an electrically powered orbital transfer vehicle

Description: To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.
Date: January 1, 1987
Creator: Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R. et al.
Partner: UNT Libraries Government Documents Department

Nuclear powered Mars cargo transport mission utilizing advanced ion propulsion

Description: Nuclear-powered ion propulsion technology was combined with detailed trajectory analysis to determine propulsion system and trajectory options for an unmanned cargo mission to Mars in support of manned Mars missions. A total of 96 mission scenarios were identified by combining two power levels, two propellants, four values of specific impulse per propellant, three starting altitudes, and two starting velocities. Sixty of these scenarios were selected for a detailed trajectory analysis; a complete propulsion system study was then conducted for 20 of these trajectories. Trip times ranged from 344 days for a xenon propulsion system operating at 300 kW total power and starting from lunar orbit with escape velocity, to 770 days for an argon propulsion system operating at 300 kW total power and starting from nuclear start orbit with circular velocity. Trip times for the 3 MW cases studied ranged from 356 to 413 days. Payload masses ranged from 5700 to 12,300 kg for the 300 kW power level, and from 72,200 to 81,500 kg for the 3 MW power level.
Date: January 1, 1987
Creator: Galecki, D.L. & Patterson, M.J.
Partner: UNT Libraries Government Documents Department

GES (Ground Engineering System) test site preparation

Description: Activities are under way at Hanford to convert the 309 containment building and its associated service wing to a nuclear test facility for the Ground Engineering System (GES) test. Conceptual design is about 80% complete, encompassing facility modifications, a secondary heat transport system, a large vacuum system, a test article cell and handing system, control and data handling systems, and safety andl auxiliary systems. The design makes extensive use of existing equipment to minimize technical risk and cost. Refurbishment of this equipment is 25% complete. Cleanout of some 1000 m/sup 3/ of equipment from the earlier reactor test in the facility is 85% complete. An Environmental Assessment was prepared and revised to incorporate Department of Energy (DOE) comments. It is now in the DOE approval chain, where a Finding of No Significant Impact is expected. During the next year, definite design will be well advanced, long-lead procurements will be initiated, construction planning will be completed, an operator training plan will be prepared, and the site (preliminary) safety analysis report will be drafted.
Date: October 1, 1987
Creator: Cox, C.M.; Mahaffey, M.K.; Miller, W.C.; Schade, A.R. & Toyoda, K.G.
Partner: UNT Libraries Government Documents Department

Lessons learned from hydrogen generation and burning during the TMI-2 event

Description: This document summarizes what has been learned from generation of hydrogen in the reactor core and the hydrogen burn that occurred in the containment building of the Three Mile Island Unit No. 2 (TMI-2) nuclear power plant on March 28, 1979. During the TMI-2 loss-of-coolant accident (LOCA), a large quantity of hydrogen was generated by a zirconium-water reaction. The hydrogen burn that occurred 9 h and 50 min after the initiation of the TMI-2 accident went essentially unnoticed for the first few days. Even through the burn increased the containment gas temperature and pressure to 1200/sup 0/F (650/sup 0/C) and 29 lb/in/sup 2/ (200 kPa) gage, there was no serious threat to the containment building. The processes, rates, and quantities of hydrogen gas generated and removed during and following the LOCA are described in this report. In addition, the methods which were used to define the conditions that existed in the containment building before, during, and after the hydrogen burn are described. The results of data evaluations and engineering calculations are presented to show the pressure and temperature histories of the atmosphere in various containment segments during and after the burn. Material and equipment in reactor containment buildings can be protected from burn damage by the use of relatively simple enclosures or insulation.
Date: May 1, 1987
Creator: Henrie, J.O. & Postma, A.K.
Partner: UNT Libraries Government Documents Department

A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

Description: This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)
Date: January 1, 1987
Creator: Bartram, B.W. & Dougherty, D.K.
Partner: UNT Libraries Government Documents Department

An analysis of loss of offsite power with a PWR at shutdown

Description: In many Probabilistic Risk Assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a loss of offsite power event that occurs while a PWR is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal capability during an outage. 5 refs., 1 fig.
Date: June 1, 1987
Creator: Chu, T.L.; Yoon, W.H. & Fitzpatrick, R.G.
Partner: UNT Libraries Government Documents Department

Evaluation of uncertainties in irradiated hardware characterization: Final report, September 30, 1986-March 31, 1987

Description: Waste Management Group, Inc. has evaluated the techniques used by industry to characterize and classify irradiated hardware components for disposal. This report describes the current practices used to characterize the radionuclide content of hardware components, identifies the uncertainties associated with the techniques and practices considered, and recommends areas for improvement which could reduce uncertainty. Industry uses two different characterization methods. The first uses a combination of gamma scanning, direct sampling, underwater radiation profiling and radiochemical analysis to determine radionuclide content, while the second uses a form of activation analysis in conjunction with underwater radiation profiling. Both methods employ the determination of Cobalt 60 content, and the determination of scaling factors for hard-to-detect Part 61 radionuclides. The accurate determination of Cobalt-60 is critical since the Part 61 activation product radionuclides which affect Part 61 classification are scaled from Cobalt-60. Current uncertainties in Cobalt-60 determination can be reduced by improving underwater radiation profiling equipment and techniques. The calculational techniques used for activation analysis can also be refined to reduce the uncertainties with Cobalt-60 determination. 33 refs., 11 figs., 10 tabs.
Date: October 1, 1987
Creator: Bedore, N.; Levin, A. & Tuite, P.
Partner: UNT Libraries Government Documents Department

Summary of HSST wide-plate crack-arrest tests and analyses

Description: Eleven wide-plate crack-arrest tests have been completed to date, seven utilizing specimens fabricated from A533B class 1 material (WP-1 series), and four fabricated from a low upper-shelf base material (WP-2 series). With the exception of one test in the WP-1 series and two tests in the WP-2 series which utilized 152-mm-thick specimens, each test utilized a single-edge notched (SEN) plate specimen 1 by 1 by 0.1 m that was subjected to a linear thermal gradient along the plane of crack propagation. Test results exhibit an increase in crack-arrest toughness with temperature, with the rate of increase becoming greater as the temperature increases. When the wide-plate test results are combined with other large-specimen results (Japanese ESSO, thermal-shock experiments and pressurized-thermal-shock experiments), the data show a consistent trend in which the K/sub Ia/ data extends above the limit provided in ASME Section XI.
Date: January 1, 1987
Creator: Naus, D.J.; Bass, B.R.; Keeney-Walker, J.; Fields, R.J.; deWit, R. & Low, S.R. III
Partner: UNT Libraries Government Documents Department

Initial pipe break analyses for advanced LMR (liquid metal reactor) concepts using MINET

Description: In support of an initial NRC review of DOE sponsored advanced liquid metal reactors (LMRs), BNL has performed some very conservative calculations of postulated primary loop pipe breaks using the MINET Code. This report briefly describes the results obtained from these calculations. 5 refs., 2 figs.
Date: June 1, 1987
Creator: Van Tuyle, G.J.; Chan, B.C. & Slovik, G.C.
Partner: UNT Libraries Government Documents Department

Concept for a small, colocated fuel cycle facility for oxide breeder fuels

Description: As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years.
Date: January 1, 1987
Creator: Burch, W.D.; Stradley, J.G. & Lerch, R.E.
Partner: UNT Libraries Government Documents Department

In-vessel flow characterization under severe accident conditions

Description: The purpose of this study is to provide a parametric framework for characterization of flow and heat transfer regimes and their associated phenomenological uncertainties following severe accidents using a two dimensional, heterogenous, porous media formulation. This approach extends the understanding of buoyancy-induced flow characteristics in the uncovered region of the reactor core and the upper plenum of a PWR vessel. The results of this study can be used to augment the boil-off steam flow in integrated one-dimensional severe accident codes such as the Source Team Code Package (STCP).
Date: January 1, 1987
Creator: Nourbakhsh, H.P.; Kim, S.B. & Khatib-Rahbar, M.
Partner: UNT Libraries Government Documents Department

Gas erosion of impeller housing in the operation of a high-temperature, high-pressure helium circulator

Description: Three gas-bearing circulators are installed in series in a high-pressure, high-temperature loop to provide helium flow up to 0.47 m/sup 3//s at a total head of 78 kJ/kg. The design pressure is 10.7 MPa, and temperatures of 1000/sup 0/C can be obtained in the test section. The inlet temperature to the circulators is limited to 450/sup 0/C. During a routine examination of the circulator, deep V-shaped grooves were found in the stationary surface of this cavity. At the same time, a very fine, dark particulate was observed in crevices of the housing. At first it was assumed that the grooves were formed by particulate erosion; however, examination of the grooves and discussions with persons experienced with large circulator operation changed this opinion. Erosion caused by particulate is characteristically rounded on the bottom and has a greater width to depth aspect than the V-shaped grooves, which were observed. Analysis of the particulate indicated that it was essentially the material of the housing that had undergone reactions with impurities in the circulating gas. It was subsequently concluded that the impeller housing had not been heat treated in a sufficiently oxidizing atmosphere after machining to form an adherent oxide coating. This suboxide coating was eroded by the shear forces in the gas. The exposed layer of metal was then further oxidized by the impurities in the gas, and these layers of oxide were successively eroded to produce the grooves. This erosion problem was eliminated by machining a ring of the same material, heat treating it to form an adherent stable oxide, and bolting it in place in the cavity.
Date: January 1, 1987
Creator: Sanders, J.P.; Heestand, R.L. & Young, H.C.
Partner: UNT Libraries Government Documents Department

Generic safety insights for inspection of boiling water reactors

Description: As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important. What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.
Date: January 1, 1987
Creator: Higgins, J.C.; Taylor, J.H.; Fresco, A.N. & Hillman, B.M.
Partner: UNT Libraries Government Documents Department

Protective action alternatives for accidents at nuclear power plants

Description: Protective action calculations have been performed for five different light water reactors (LWRs) and containment designs using high and low fission product releases for early and late containment failures for each plant. These fission product release estimates were obtained from studies performed for the recently published ''Reactor Risk Reference Document'' (NUREG-1150). Five protective actions were considered for the risks of exceeding various dose levels to the red marrow versus centerline distance from the plants using site-specific meteorology. The strategies considered were 4 hours of normal activity, basement sheltering, large building sheltering, evacuation at release, and evacuation 1 hour after release. The evacuations were computed using 10 mph evacuation speed for all sites. Additional calculations were performed for the dose contributions due to the cloud, ground, and inhalation pathways.
Date: June 1, 1987
Creator: Tingle, A.; Pratt, W.T. & McGuire, S.A.
Partner: UNT Libraries Government Documents Department

Handbook of software quality assurance techniques applicable to the nuclear industry

Description: Pacific Northwest Laboratory is conducting a research project to recommend good engineering practices in the application of 10 CFR 50, Appendix B requirements to assure quality in the development and use of computer software for the design and operation of nuclear power plants for NRC and industry. This handbook defines the content of a software quality assurance program by enumerating the techniques applicable. Definitions, descriptions, and references where further information may be obtained are provided for each topic.
Date: August 1, 1987
Creator: Bryant, J.L. & Wilburn, N.P.
Partner: UNT Libraries Government Documents Department

Impact of Zr metal and coking reactions on the ex-vessel source term predictions of CORCON/VANESA

Description: During a core meltdown accident in a LWR, molten core materials (corium) could leave the reactor vessel and interact with concrete. In this paper, the impact of the zirconium content of the corium pool and the coking reaction on the release of fission products are quantified using CORCON/Mod2 and VANESA computer codes. Detailed calculations show that the total aerosol generation is proportional to the zirconium content of the corium pool. Among the twelve fission product groups treated by the VANESA code, CsI, Cs/sub 2/O and Nb/sub 2/O/sub 5/ are completely released over the course of the core/concrete interaction, while an insignificant quantity of Mo, Ru and ZrO/sub 2/ are predicted to be released. The release of BaO, SrO and Ce/sub 2/O increase, while the releases of Te and La/sub 2/O/sub 3/ are relatively unaffected by the Zr content of the corium pool. The impact of the coking reaction on the radionuclide release and aerosol production was found to be insignificant.
Date: January 1, 1987
Creator: Lee, M.; Davis, R.E. & Khatib-Rahbar, M.
Partner: UNT Libraries Government Documents Department