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UPDATE: nuclear power program information and data, July-September 1981

Description: UPDATE is published by the Office of Coordination and Special Projects, Office of Nuclear Reactor Programs, to provide a quick reference source on the current status of nuclear powerplant construction and operation in the United States and for information on the fuel cycle, economics, and performance of nuclear generating units. Similar information on other means of electric generation as related to nuclear power is included when appropriate. The subject matter of the reports and analyses presented in UPDATE will vary from issue to issue, reflecting changes in foci of interest and new developments in the field of commercial nuclear power generation. UPDATA is intended to provide a timely source of current statistics, results of analyses, and programmatic information proceeding from the activities of the Office of Nuclear Reactor Programs and other components of the Department of Energy, as well as condensations of topical articles from other sources of interest to the nuclear community. It also facilitates quick responses to requests for data and information of the type often solicited from this office.
Date: January 1, 1981
Creator: /NBM--6011986, DOE
Partner: UNT Libraries Government Documents Department

Computer simulation of LMFBR piping systems. [Accident conditions]

Description: Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses.
Date: January 1, 1977
Creator: A-Moneim, M.T.; Chang, Y.W. & Fistedis, S.H.
Partner: UNT Libraries Government Documents Department

Markov Model of Accident Progression at Fukushima Daiichi

Description: On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.
Date: November 11, 2012
Creator: A., Cuadra; R., Bari; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G. et al.
Partner: UNT Libraries Government Documents Department

Confinement of airborne radioactivity. Final progress report, January-December 1978

Description: A new test method has been developed at the Savannah River Laboratory for evaluating the iodine retention capabilities of carbon used in the airborne-activity confinement system. Methyl iodide tagged with I-131 is injected into a test gas stream continuously for 5 hours with test conditions of 80/sup 0/C temperature, 95% relative humidity, and 55 feet per minute linear flow velocity. Results show that the CH/sub 3/I retention efficiency is independent of the inlet CH/sub 3/I concentration over the range of at least 0.9 to 200 ..mu..g/m/sup 3/ in the test gas stream. The method was also used to evaluate the effects of paint fumes on in-service carbons and showed that solvent exposure reduced carbon service life by 5 to 7 months. Experimental carbons both before and after service exposure in the SRP carbon test facility were also evaluated.
Date: February 1, 1980
Creator: A.G., Evans
Partner: UNT Libraries Government Documents Department

CONFIRMATORY SURVEY OF THE FUEL OIL TANK AREA HUMBOLDT BAY POWER PLANT EUREKA, CALIFORNIA

Description: During the period of February 14 to 15, 2012, ORISE performed radiological confirmatory survey activities for the former Fuel Oil Tank Area (FOTA) and additional radiological surveys of portions of the Humboldt Bay Power Plant site in Eureka, California. The radiological survey results demonstrate that residual surface soil contamination was not present significantly above background levels within the FOTA. Therefore, it is ORISE’s opinion that the radiological conditions for the FOTA surveyed by ORISE are commensurate with the site release criteria for final status surveys as specified in PG&E’s Characterization Survey Planning Worksheet. In addition, the confirmatory results indicated that the ORISE FOTA survey unit Cs-137 mean concentrations results compared favorably with the PG&E FOTA Cs-137 mean concentration results, as determined by ORISE from the PG&E characterization data. The interlaboratory comparison analyses of the three soil samples analyzed by PG&E’s onsite laboratory and the ORISE laboratory indicated good agreement for the sample results and provided confidence in the PG&E analytical procedures and final status survey soil sample data reporting.
Date: April 9, 2012
Creator: ADAMS, WADE C.
Partner: UNT Libraries Government Documents Department

Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

Description: A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated.
Date: December 1, 1979
Creator: Abbott, L.S.; Bartine, D.E. & Burns, T.J. (eds.)
Partner: UNT Libraries Government Documents Department

Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

Description: A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.
Date: December 1, 1978
Creator: Abbott, L.S.; Bartine, D.E. & Burns, T.J. (eds.)
Partner: UNT Libraries Government Documents Department

Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

Description: During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.
Date: January 1, 1982
Creator: Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J. & Slater, C.O.
Partner: UNT Libraries Government Documents Department

AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

Description: This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).
Date: August 1, 2011
Creator: Abbott, Michael L. & Daum, Keith A.
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Part II. A. Pool growth in composite beds; B. Effect of overlaying steel layers. Final report, March 1, 1978-September 30, 1979. [LMFBR]

Description: The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid.
Date: December 26, 1979
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Progress report, April 1--June 30, 1979. [LMFBR]

Description: The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas-releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid. At this time pool inversion occurs and the immiscible liquid sinks to the bottom of the pool displacing the now lighter fuel-concrete simulant.
Date: July 1, 1979
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

Growth of molten core debris pools in concrete. Progress report, March 1, 1977--November 30, 1977. [LMFBR]

Description: Experiments have been conducted using a volumetrically-heated pool with noncondensable gas injection at the boundaries to simulate the heat transfer processes taking place in molten core debris/concrete systems. Measurements of the upward, downward, and sideward heat transfer rates have been made over wide ranges of internal Rayleigh number (6.0 x 10/sup 5/ < Ra < 2.7 x 10/sup 9/), superficial gas velocity (0 < V/sub g/ < 2.5 cm/min), and pool aspect ratio (1.32 < W/L < 5.16). Pools with either an isothermal solid upper boundary or a free surface have been examined. Detailed measurements of the temperature distribution within the pools have also been made. The results indicate that the downward heat transfer rates are significantly increased while the sideward heat transfer rates are slightly affected by gas injection. The heat transfer rates at the pool boundaries are sensitive to the upper surface boundary condition. Efforts are continuing on the second part of the present study where the phase-change and gas evolution processes are combined.
Date: November 1, 1977
Creator: Abdel-Khalik, S I
Partner: UNT Libraries Government Documents Department

Debris-bed cooling following an HCDA in a fast reactor. Annual progress report

Description: Natural convection within simulated core debris beds has been investigated experimentally and numerically. The objectives of this research have been: (1) to develop a three-dimensional transient model for analyzing single-phase cooling of debris beds; (2) to validate the model using out-of-pile simulant experiments which measure the detailed structure of the convection cells within the bed as well as integral heat transfer rates; and (3) to apply the model to typical core debris beds over a wide range of parameters in order to determine the relative importance of conduction and convection in sodium-cooled debris beds. The bed has been simulated using directly heated two-dimensional packed tube bundles of different particle diameters, bed loadings, porosities, heat generation rates, and overlaying fluid layer heights. Detailed velocity and temperature profiles within the bed and overlaying fluid have been measured using particle tracing techniques and Mach-Zehnder interferometry. Measurements of the downward and upward power fractions and Nusselt numbers over a wide range of experimental variables have been made. The data have been compared with predictions of the transient three-dimensional natural convection computer code COMMIX-1A. 14 figures.
Date: October 1, 1982
Creator: Abdel-Khalik, S.I.
Partner: UNT Libraries Government Documents Department

Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

Description: Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.
Date: July 2, 2005
Creator: Abdel-Khalik, Said
Partner: UNT Libraries Government Documents Department

Neutronic Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 - Volume 4, Part 2--Saxton Plutonium Program Critical Experiments

Description: Critical experiments with water-moderated, single-region PuO{sub 2}-UO{sub 2} or UO{sub 2}, and multiple-region PuO{sub 2}-UO{sub 2}- and UO{sub 2}-fueled cores were performed at the CRX reactor critical facility at the Westinghouse Reactor Evaluation Center (WREC) at Waltz Mill, Pennsylvania in 1965 [1]. These critical experiments were part of the Saxton Plutonium Program. The mixed oxide (MOX) fuel used in these critical experiments and then loaded in the Saxton reactor contained 6.6 wt% PuO{sub 2} in a mixture of PuO{sub 2} and natural UO{sub 2}. The Pu metal had the following isotopic mass percentages: 90.50% {sup 239}Pu; 8.57% {sup 239}Pu; 0.89% {sup 240}Pu; and 0.04% {sup 241}Pu. The purpose of these critical experiments was to verify the nuclear design of Saxton partial plutonium cores while obtaining parameters of fundamental significance such as buckling, control rod worth, soluble poison worth, flux, power peaking, relative pin power, and power sharing factors of MOX and UO{sub 2} lattices. For comparison purposes, the core was also loaded with uranium dioxide fuel rods only. This series is covered by experiments beginning with the designation SX.
Date: October 12, 2000
Creator: Abdurrahman, NM
Partner: UNT Libraries Government Documents Department

Speciation studies of radionuclides in low level wastes and process waters from pressurized water reactor

Description: Physicochemical characterization studies of aqueous process streams at San Onofre Generating Station Unit No. 1 are providing important source term information concerning the forms of radionuclides being released into the marine environment. The primary coolant, secondary steam condensate, processed low level wastes and tertiary coolant were sampled at several different times in the nuclear fuel cycle. Rdionuclides were partitioned into particulate, cationic, anionic, and nonionic species in the reactor process streams, and into particulate and soluble species in the tertiary seawater coolant. Characterization of the particulate species has included a detailed size distribution. The purpose of this research was to provide information concerning chemical and physical forms of the radionuclides being released to the coastal zone from a nuclear generating station in order to facilitate design of radiobiological studies necessary for assessment of their environmental significance.
Date: October 1, 1977
Creator: Abel, K. H.; Robertson, D. E.; Crecelius, E. A. & Silker, W. B.
Partner: UNT Libraries Government Documents Department

Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations

Description: A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research program examined radionuclide concentrations and distributions external to the reactor vessel at seven US light water reactors. Later stages of the research program have examined the radionuclide distributions in the highly radioactive reactor internals and fuel assembly. Most recently, the research program is determining radionuclide concentrations in these highly radioactive components and comparing empirical results with those derived from the several nonempirical methodologies employed to estimate radionuclide inventories for disposal classification. The results of the research program to date are summarized, and their implications and significance for the decommissioning process are noted.
Date: September 1, 1992
Creator: Abel, K.H.; Robertson, D.E. & Thomas, C.W.
Partner: UNT Libraries Government Documents Department

Radioactive waste management at a Liquid Metal Fast Breeder Reactor

Description: This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel.
Date: January 1, 1979
Creator: Abrams, C.S.; Fryer, R.H. & Witbeck, L.C.
Partner: UNT Libraries Government Documents Department

Calculational limitations in PWR system simulation

Description: Engineering transient analysis codes, which are in general more accurate than the present generation of simulator software, can be expected to yield reasonably accurate results (+-20% or so on system pressure) if carefully utilized and if the two-phase and transient flow conditions are not severe. As the severity of the transient increases, the confidence that one may have in the results decreases. None of the existing engineering analysis codes is well assessed or verified for transient analysis, but all give qualitatively the same results lending credence to their results. Recent comparisons to transients in LOFT and SEMISCALE are encouraging as are various comparisons to actual plant data.
Date: January 1, 1982
Creator: Abramson, P.B.; Kennedy, M.F. & Speis, T.P.
Partner: UNT Libraries Government Documents Department

Optical probe for local void fraction and interface velocity measurements. [BWR; PWR]

Description: In view of the importance of obtaining unsteady local void fraction and interface velocities in liquid-vapor two-phase flows, an optical probe with a controlled tip geometry was developed and is described. In order to minimize the disturbances caused to the flow field by the presence of the probe, its dimensions have been miniaturized. The electronic and hydrodynamic response of the probe were investigated experimentally. The probe was found to be sensitive to both the interface velocities and the phase present at the probe tip. A possible explanation for the behavior of the probe is presented. Within the velocity range checked and with proper calibration, the optical probe developed can be used to determine both local void fractions and interface velocities.
Date: March 1, 1978
Creator: Abuaf, N.; Jones, O.C. Jr. & Zimmer, G.A.
Partner: UNT Libraries Government Documents Department

PWR AXIAL BURNUP PROFILE ANALYSIS

Description: The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).
Date: September 17, 2003
Creator: Acaglione, J.M.
Partner: UNT Libraries Government Documents Department

Tritium distribution in the MHTGR

Description: The {sup 3}H production, transport and environmental release from the 350 MW(t) Modular High Temperature Gas Cooled Reactor was analyzed. The analysis was performed using a modified TRITGO computer code, plant data base from the Preliminary Safety Information Document and materials property data from the Fuel Design Data Manual, Issue F. The analysis indicates that most of the {sup 3}H produced in the reactor is retained by the fuel particles and the structural graphite elements. The single largest source of {sup 3}H is ternary fission in the fuel particles, of which 95% is retained by the particles. The {sup 3}H released from the core and the {sup 3}H produced by {sup 3}He activation are largely removed by the Helium Purification System. Assuming zero leakage of water from the secondary system, the average predicted {sup 3}H activity in the secondary water of 0.35 {mu}Ci/g is much greater than the allowable activity of 5 pCi/g for direct discharge into the environment. If any of the secondary water has to be discharged, it must be diluted prior to discharge. 10 refs., 9 figs.
Date: May 1, 1988
Creator: Acharya, R.
Partner: UNT Libraries Government Documents Department

Oxidation of 316 stainless steel and other alloys in prototypic GCFR environments

Description: The oxidation behavior of type 316 stainless steel and candidate advanced alloys for the gas-cooled fast reactor (GCFR) is being investigated at General Atomic Company. The test program consists of oxidation tests in prototypic GCFR environments. Two tests have been completed to date and a third test is under way. The first test was performed in an environment containing a hydrogen/water ratio of 10. The oxidation behavior of all the alloys was good to excellent in this environment. Preferential oxidation of chromium was responsible for this behavior. The second test was performed in an environment containing a hydrogen/water ratio of 0.25, where both chromium and iron oxides are thermodynamically stable. Some of the alloys and some of the ribbed type 316 stainless steel test specimens showed unacceptable oxidation resistance in this environment. In the third test, presently under way, two different pretreatment procedures are being used to control the poor oxidation behavior observed in the second test. Early results show some degree of success.
Date: May 1, 1980
Creator: Acharya, R.T.
Partner: UNT Libraries Government Documents Department

Gas-cooled fast breeder reactor steady-state irradiation testing program

Description: The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.
Date: August 1, 1980
Creator: Acharya, R.T.; Campana, R.J. & Langer, S.
Partner: UNT Libraries Government Documents Department