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Pressurized Water Reactor Program Technical Progress Report for the Period July 15, 1954 to August 26, 1954

Description: This progress report has 2 parts. Part 1 PWR Engineering covers the following topics: (1) power plant analysis and systems; (2) power plant components and component materials and tests; and (3) reactor and auxiliaries. Part 2 PWR Development covers: (1) fuel element development; (2) metallurgy of core materials; (3) materials application development; (4) chemistry development; (5) irradiation effects; and (6) reactor physics.
Date: October 31, 1957
Partner: UNT Libraries Government Documents Department

Presentation to the Atomic Energy Commission and the Air Force, June 14, 1962

Description: This volume contains the charts and backup material presented to the Atomic Energy Commission and Air Force on June 14, 1962 concerning General Electric's Nuclear Materials and Propulsion Operation (formerly the Aircraft Nuclear Propulsion Department), during its work on the development of a nuclear power plant for manned aircraft.
Date: October 1, 1962
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY. Design and development quality assurance requirements for the FFTF

Description: The document is presented to provide general management requirements for Pacific Northwest Laboratory (PNL) and contractor design and development quality assurance programs to assure the required quality level of the various items required for the FFTF. The document is applicable as imposed by the contract to FFTF contractors and subcontractors. The document is also applicable to PNL design and development activities related to the FFTF.
Date: October 23, 1968
Creator: Albert, W.G.
Partner: UNT Libraries Government Documents Department

Analysis of homogeneous U233 and U235 critical assemblies with ENDF/B-IV data (AWBA development program)

Description: Thirty-two U233 and U235 homogeneous aqueous critical experiments were analyzed with ENDF/B-IV data. Calculated eigenvalues for both fuel types increased by nearly 2 percent over the range of hydrogen/uranium atomic ratio covered (from 2106 to 27.1). This is attributed mostly to an underprediction of fast leakage, with some contribution from the fission and capture resonance integrals of ENDF/B-IV U235. Eigenvalue sensitivities to several nuclear data changes were examined. Values of the thermal criticality parameter constraint K2 for U233 and U235 were derived from the Gwin-Magnuson critical experiments at the zero leakage limit.
Date: October 1, 1977
Creator: Ullo, J.J. & Hardy, J. Jr.
Partner: UNT Libraries Government Documents Department

Behavior of fission product gases in HTGR fuel material

Description: This document is a status report on experimental data relating to the behavior of fission product gases in HTGR fuel material. It contains (1) a review of published experimental data on the behavior of gaseous fission products (noble gases, halogens, and related elements) in HTGR fuel materials and (2) new data on fission gas behavior. From the experimental data, the effects of fission product element, fuel configuration, nuclide half-life, temperature, pressure, neutron flux, neutron fluence, and fuel hydrolysis on fission gas release are deduced. Also on the basis of the experimental data, a model is formulated for the release of fission gases from fuel rods. The data presented in this report are being confirmed and extended through continuing work at General Atomic and other laboratories.
Date: October 1, 1977
Creator: Myers, B.F.; Baldwin, N.L.; Bell, W.E. & Burnette, R.D.
Partner: UNT Libraries Government Documents Department

Near-isotropic petroleum-coke based graphites for high temperature gas-cooled reactor core components

Description: The standard covers procurement requirements for extruded graphite logs, 15 in. (381 mm) or greater in diameter, manufactured with near-isotropic petroleum cokes and coal-tar pitch binders which are candidates or reference materials for replaceable fuel and reflector blocks for High-Temperature Gas-Cooled Reactors (HTGRs). The requirements are designed to produce the degree of lot-to-lot reproducibility which is required to ensure consistent and predictable properties and irradiation performance for specific graphite grades and to ensure traceability of the graphite logs to production processes and raw materials that affect performance. The standard is intended for use in the procurement of developmental and commercial grades of nuclear graphite which are to be evaluated on Department of Energy (DOE) funded programs for use as core components in HTGRs.
Date: October 1, 1977
Partner: UNT Libraries Government Documents Department

Posttest REALP4 analysis of LOFT experiment L1-3A

Description: This report presents selected results of posttest RELAP4 modeling of LOFT loss-of-coolant experiment L1-3A, a double-ended isothermal cold leg break with lower plenum emergency core coolant injection. Comparisons are presented between the pretest prediction, the posttest analysis, and the experimental data. It is concluded that pressurizer modeling is important for accurately predicting system behavior during the initial portion of saturated blowdown. Using measured initial conditions rather than nominal specified initial conditions did not influence the system model results significantly. Using finer nodalization in the reactor vessel improved the prediction of the system pressure history by minimizing steam condensation effects. Unequal steam condensation between the downcomer and core volumes appear to cause the manometer oscillations observed in both the pretest and posttest RELAP4 analysis.
Date: October 1, 1977
Creator: White, J.R. & Holmstrom, H.L.O.
Partner: UNT Libraries Government Documents Department

Review of FFTF and CRBRP control rod systems designs

Description: 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.
Date: October 4, 1977
Creator: Pitterle, T. A. & Lagally, H. O.
Partner: UNT Libraries Government Documents Department

LOFT Steam Generator thermal analysis Class I review

Description: A Class 1 review of the thermal analysis of the LOFT Steam Generator and of the transient thermal analysis of LOFT steam generator thermal well was made to confirm compliance of the analyses with the LOFT Technical Specifications, with the LOFT integral test system - preliminary component design description for the steam generator, and with Appendix C of ANC Specification 60139. Included in this review is an examination of the thermal transients and conditions to insure adequate conservatism in analyzed conditions and transients to which the LOFT Steam Generator (including thermal well) might be exposed while in service at the LOFT reactor facility. Also, a review of the thermal code used was made to check for its applicability to these analyses.
Date: October 6, 1977
Creator: Kinnaman, T.L.
Partner: UNT Libraries Government Documents Department

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

Description: An accurate Mark I /sup 1///sub 5/-scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data.
Date: October 11, 1977
Creator: Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K. & McCauley, E.W.
Partner: UNT Libraries Government Documents Department

Sampling and characterization of sodium-water reaction products

Description: Sodium-water reaction products (SWRP) which had accumulated in the bottom of the reaction products tank (RPT) of the Large Leak Test Rig (LLTR) were sampled and characterized. Analysis showed that the SWRP consisted of NaOH, Na/sub 2/O, NaH, and Na, as expected, in varying proportions. The unreacted sodium in the samples examined ranged from 32 to 60% by weight. The SWRP reacts (dissolves) rapidly and completely with ethanol, and somewhat less rapidly and completely with Dowanol PM. A magnetic, metallic residue was left, originating in the Croloy of the test article in the LLTR. These solvents should be investigated further as possible agents for cleaning sodium systems that are highly contaminated with water reaction products. The SWRP does not completely liquefy with heating to 800/sup 0/F (427/sup 0/C). Although it softens to a consistency similar to that of wet sand, it is unlikely that it would flow or could be pumped even at such elevated temperature. The presence of highly corrosive molten NaOH. (above 606/sup 0/F, 319/sup 0/C) also makes the hot draining of SWRP unattractive.
Date: October 13, 1977
Creator: Eichelberger, R.L.
Partner: UNT Libraries Government Documents Department

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

Description: The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and development of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.
Date: October 1, 1979
Creator: None
Partner: UNT Libraries Government Documents Department

Application of Linear Propagation of Errors to Fuel Rod Temperature and Stored Energy Calculations

Description: Linear propagatlon of errors evaluates modeling uncertainty by approximating a function of interest by first-order Taylor's series expansions and then approximating the variance of the function by the variance of the linear approximation. This report discusses uncertainty analysis for different nuclear fuel rod designs, the process of model validation, and the effect of cracked pellet fuel models upon temperabre uncertainty. Using a postulated power history, the uncertainty for the predicted thermal response of boiling water reactor (BWR) and pressurized water reactor (PWR} fuel rods was evaluated. Beginning-of-life (BOL) relative uncertainty for BWR and PWR fuel rods is approximately the same. while different end-of-fife {EOL} thermal response results in different EOL uncertainty. Determining the validity of modeling relative to reality is discussed in qualitative terms. Validity is dependent upon verifying that the code correctly implements the model and that satisfactory agreement is found between the model and measurements. Fuel modeling codes are now using cracked pellet fuel models, which result in decreased fuel surface temperature. Estimated stored energy is lowered; but its relative uncertainty is increased. In general, however, the absolute upper uncertainty bound for stored energy is lower for a cracked pellet model than for a solid pellet model.
Date: October 1, 1980
Creator: Cunningham, M. E.; Olsen, A. R.; Lanning, D. D. & Willford, R. E.
Partner: UNT Libraries Government Documents Department

Final report of comprehensive testing program for concrete at elevated temperatures

Description: The objective of this program was to define the variations in physical (thermal) and mechanical (strength) properties of limestone aggregate concrete and lightweight insulating concrete exposed to elevated temperatures that could occur as a result of a postulated large sodium spill in a lined LMFBR equipment cell. To meet this objective, five test series were conducted: (1) unconfined compression, (2) shear, (3) rebar bond, (4) sustained loading (creep), and (5) thermal properties. Mechanical property results are presented for concretes subjected to temperature up to 621{sup 0}C (1150{sup 0}F).
Date: October 1, 1980
Creator: Oland, C.B.; Naus, D.J. & Robinson, G.C.
Partner: UNT Libraries Government Documents Department

LOFT Monthly Progress Report for September 1980

Description: The fourth nuclear powered small break test (L3-5/5A) was conducted on September 29, 1980. The test was initiated from a steady state operating condition wherein the core was generating heat at a maximum rate of approximately 52 kW/m. The test consisted of two parts: L3-5 simulated a 4-in. pipe break in a commerical pressurized water reactor; the second part, L3-5A, was intended to investigate natural circulation and steam generator heat transfer modes and also plan recovery using secondary system control in a situation where the pipe break and the ECCS accumulator are isolated from the primary coolant system. Initial test data indicated that all systems functioned as expected. The several hundred measurements of system coolant and reactor core conditions made during the three hour duration of the test will continue to be analyzed over the next several months. Preparations were also underway for conduting three tests in the Anticipated Transient Series. These tests, designated L6-1, L6-2, and L6-3 will provide information on plant control systems and operator response to transients in which the initiating event is not a loss-of-primary coolant. The transient tests to be conducted during September and others scheduled in the future will add greatly to understanding responses necessary to a transient condition. September 1980 marked the successful completion of FY-1980. Final closing values for each of the funding sources are included as part of this report. NRC and foreign funded tasks closed FY-1980 with underruns documented by identifying committed and uncommitted carryovers.
Date: October 1, 1980
Creator: Kaufman, N. C.
Partner: UNT Libraries Government Documents Department

Population Dose Commitments Due to Radioactive Releases from Nuclear Power Plant Sites in 1977

Description: Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1977. Fifty-year dose commitments from a one-year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ, Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 220 person-rem to a low of 0.003 person-rem with an arithmetic mean of 16 person-rem. The total population dose for all sites was estimated at 700 person-rem for the 92 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 2 x 10{sup -5} mrem to a high of 0.1 mrem. No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites.
Date: October 1, 1980
Creator: Baker, D. A.
Partner: UNT Libraries Government Documents Department

Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization

Description: This report summarizes work performed for the U.S. Arms Control and Disarmament Agency under BOA AC9NX707 (Task Order 80-02), as part of the Agency's continuing program on improved fuel utilization in light water reactors. The objective of the study was to investigate improvements in fuel management and design of water reactors (PWRs) that could potentially increase the utilization of natural uranium resources in a once-through fuel cycle (i.e., without using spent fuel reprocessing and recycle). For the present study, potential improvements were limited to retrofittable concepts, i.e., those which could be modifications to the reactor system or balance of plant. The potential improvements considered were not necessarily restricted to those which might be economical under current uranium ore prices or to those which might be acceptable to the nuclear industry at the present time. A six-month fuel cycle, for example, although technically possible, would be neither economical nor accept able to the industry at the present time. Although all potential improvements are not necessarily compatible with each other, the target objective was to seek a composite system of compatible improvements that, if possible, could increase uranium resource utilization by 30% or more. Economic factors, risks involved in the introduction, and potential licensing concerns are also addressed in the report.
Date: October 1, 1980
Partner: UNT Libraries Government Documents Department

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

Description: Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable.
Date: October 1, 1980
Creator: Ryan, J. A. & Mahoney, J. J.
Partner: UNT Libraries Government Documents Department

Some aspects of materials development for sodium heated steam generators

Description: A development program was undertaken to support the materials selection for steam generator piping and IHX which are to be used in Liquid Metal Fast Breeder Reactors (LMFBR). Four major topics were reviewed, describing the results obtained as well as the direction of future tests. These topics are: carbon transport in sodium, effect of carbon loss/gain upon materials in the reactor Intermediate Heat Transport System (IHTS), corrosion fatigue and aqueous corrosion. The results support the initial assumptions made in specifying the use of 2-1/4Cr-1Mo as the construction material for the evaporator and superheater and Type 316 piping of the IHT system. Future direction of the experimental programs is to further verify the materials choice and to also obtain information which will be essential during the plant installation, operation and reliability of the components.
Date: October 1, 1980
Creator: Roy, P. & Spalaris, C.N.
Partner: UNT Libraries Government Documents Department

Characterization of radioactive ion exchange media waste generated at Three Mile Island

Description: The March 1979 accident at General Public Utilities Nuclear Corporation (GPUNC) Three Mile Island Nuclear Power Station Unit 2 (TMI-2), resulted in the transfer of more than 1100 m/sup 3/ of contaminated water to the auxiliary and fuel handling building. The principal sources of the water were the makeup and letdown purification system and the containment building sump. The contaminated water was processed through an ion exchange system designated as EPICOR II. The EPICOR-II System is a three-stage process. The contaminated water passes through a first stage of ion exchange media, designated as prefilters, and then through the second and third stages, designated as demineralizers. The majority of the activity was deposited in the first-stage prefilters, which have a maximum administrative loading limit of 1300 curies. The predominant radionuclides present in the prefilters are cesium and strontium.
Date: October 1, 1981
Creator: Runion, T.C.; Holzworth, R.E.; Ogle, R.E.; Burton, H.M. & Bixby, W.W.
Partner: UNT Libraries Government Documents Department

Sodium removal process development for LMFBR fuel subassemblies

Description: Two 37-pin scale models of Clinch River Breeder Reactor Plant fuel subassemblies were designed, fabricated and used at Westinghouse Advanced Reactors Division in the development and proof-testing of a rapid water-based sodium removal process for the ORNL Hot Experimental Facility, Liquid Metal Fast Breeder Reactor Fuel Reprocessing Cycle. Through a series of development tests on one of the models, including five (5) sodium wettings and three (3) high temperature sodium removal operations, optimum process parameters for a rapid water vapor-argon-water rinse process were identified and successfully proof-tested on a second model containing argon-pressurized, sodium-corroded model fuel pins simulating the gas plenum and cladding conditions expected for spent fuel pins in full scale subassemblies. Based on extrapolations of model proof test data, preliminary process parameters for a water vapor-nitrogen-water rinse process were calculated and recommended for use in processing full scale fuel subassemblies in the Sodium Removal Facility of the Fuel Receiving Cell, ORNL HEF.
Date: October 1, 1981
Creator: Simmons, C.R. & Taylor, G.R.
Partner: UNT Libraries Government Documents Department

Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation

Description: Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements.
Date: October 1, 1985
Creator: Hanson, D.L.; Jovanovic, V. & Burnette, R.D.
Partner: UNT Libraries Government Documents Department

Testimony (on space nuclear power) before the Subcommittee on Energy Research and Production of the Committee on Science and Technology, US House of Representatives

Description: This brief presentation consists of four parts: first, a few introductory comments about space nuclear power technology; second, a description of activities currently underway at Martin Marietta Energy Systems; third, a discussion of future directions at Martin Marietta Energy Systems; and fourth, some recommendations.
Date: October 9, 1985
Creator: Mynatt, F.R.
Partner: UNT Libraries Government Documents Department

Integration of wide-plate crack-arrest test results

Description: The primary objective of the crack-arrest studies under the Heavy-Section Steel Technology (HSST) Program is to generate data for understanding the crack-arrest behavior of prototypical pressure vessel steels at temperatures near and above the onset of the Charpy uppershelf region. Specific program goals are to: (1) extend existing K/sub Ia/ data bases to temperatures beyond those associated with the upper limit in the ASME BandPVC; (2) clearly establish that crack-arrest occurs prior to fracture-mode conversion; (3) observe the relationship between arrest data and machine/specimen compliance behavior; and (4) validate the predictability of crack arrest, stable tearing, and/or unstable tearing sequences for ductile materials. In meeting these goals, the HSST program is generating crack-arrest data over an expanded temperature range through tests involving large thermally-shocked cylinders, pressurized-thermally-shocked vessels and wide-plate specimens. This presentation will focus on data from the wide-plate specimens which have the advantage that a more significant number of data points can be obtained at affordable costs.
Date: October 1, 1986
Creator: Pugh, C.E. & Naus, D.J.
Partner: UNT Libraries Government Documents Department