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CRITICAL EXPERIMENTS FOR THE PRELIMINARY DESIGN OF THE ARGONNE HIGH FLUX REACTOR
Critical experiments were performed with two assemblies simulating a cold clean, and an end-of-cycle,- Argonne High Flux Reactor, core. Data were obtained for flux distributions; cadmium ratios; temperature and void coefficients; and control rod, beam hole, and reflector worths. The data obtained furnished confirmation of theoretical predictions. The peak 2200-m/sec flux per unit power was measured as 3 x 10/sup 7/ n/(cm/sup 2/)(sec)(watt) for both cores. The two cores had internal H/sub 2/O thermal columns, 12.7 cm x 12.7 cm x 50.8 cm. These were enclosed by 100-liter fuel zones. The radial reflector was 90% beryllium containing 10% H/sub 2/0 plus Plexiglas by volume. The top and bottom reflectors were H/sub 2/O. The critical mass was 3.58 kg U/sup 235/ with a 1.16 metal-towater ratio in the fuel zone. The critical mass with a 1.60 metal- to-water ratio, taking into account 34.3 kg Type 304 stainless steel, was 7.15 kg U/sup 235/. (auth)
Date:
June 1, 1961
Creator:
de Villiers, J.W.L. ed.
Object Type:
Report
System:
The UNT Digital Library
Transient Temperature Distributions in a Thermally Orthotropic Plate With Non-Uniform Surface Heating
ent temperature variation in a thermally orthotropic plate which is subjected to an arbitrary heating rate distribution along one face with all other surfaces being insulated. Dimensionless temperature histories and distributions determined from this solution are presented for the special, but representative, case of a linearly varying heating rate distribution on plates with varying degrees of thermal orthotropy. These results establish quantitatively the value of a material with high planar and low normal thermal conductivities for applications where it is desired to maintain minimum temperatures on the rear or unheated surface of a heat shield when the heated surface is subjected to a very non-uniform heating rate distribution. The applicability of simplifying assumptions in analyzing such a system is discussed. Experimental temperature measurements in a pyrolytic graphite plate heated by an oxyacetylene flame were made to verify the analytical results. Achievement of satisfactory agreement wss found to be dependent upon use of thermal property values differing from those presently available for this material. This is not unusual in that differences in production methods are known to introduce substantial property variations in anisotropic materials such as pyrolytic graphite. (auth)
Date:
June 1, 1961
Creator:
Hornbaker, David Ross
Object Type:
Thesis or Dissertation
System:
The UNT Digital Library
A Theoretical Study of the Transient Operation and Stability of Two-Phase Natural Circulation Loops
Mathematical models of the time-dependent behavior of two-phase natural- circulation loops were used to predict the operation and to explain the unusual instability sometimes observed. The initial results obtained for a loop similar to the Univ. of Minnesota loop were used to formulate a more complex and accurate model, and the predicted transient behavior was in close agreement with the experimental results from the Minnesota loop. For a 300psia, high-pressure loop, unstable oscillatory behavior was predicted under certain conditions and stable behavior under others. Closed unstable regions rather than limits were predicted, and the specifications of stability in terms of a single parameter were found to be impossible. The great difference in oscillatory frequencies observed at low and high pressures was found to be due largely to the system geometry. The criterion for the absence of oscillations was found to be similar to one of the criteria for stability of chemical reaction systems. (D.L.C.)
Date:
June 1, 1961
Creator:
Garlid, K.; Amundson, N. R. & Isbin, H. S.
Object Type:
Thesis or Dissertation
System:
The UNT Digital Library
Periodic Waste Disposal System Material Balance Test. Core 1, Seed 2. Test Evaluation T-641317. Section 1
A test was carried out to determine the adequacy of storage capacity and operating procedures of the radioactive waste disposal system during a normal reactor plant warmup. The capacity and operating procedures were found to be adequate. It was impossible to perform a complete material balance based on existing level instrumentation and using the data required by the test procedure. Approximately 21,290 gal. of waste were received in the system and 13,210 gal. were discharged to the river with a total activity of 1200 mu c. A quantity of 6,670 gal. of reactor coolant effluent was processed. Approximately 634 lb of combustible waste were incinerated. (M.C.G.)
Date:
June 30, 1961
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
PL FINAL DESIGN REPORT. VOLUME II. PLANT DRAWINGS
Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)
Date:
June 30, 1961
Creator:
Combustion Engineering, Inc. Nuclear Div., Windsor, Conn.
Object Type:
Report
System:
The UNT Digital Library
Corrosion Tests in Molten Lead-Lead Chloride
Corrosion tests were run on some commercial grade metals, an alloy steel, stainless steels, chromium-- nickel-iron alloys, nickel base alloys, cobalt base alloys, and a chromium-- nickel-- cobalt-- iron ailoy in the system: leadlead chloride-lead chloride vapor at 528 deg C under an argon atmosphere. The following metals and alloys showed a corrosion rate of nine mils per month or less and did not suffer intergranular or other localized attack: tantalum, Incoloy 804, Hastelloy F, Carpenter-20 (Cb), stainless steels 316L, 318 Cb, Haynes Alloy 25, and Haynes Multimet (auth)
Date:
June 1, 1961
Creator:
Stolica, N. D.; Adams, G. S. & Bomar, M. R.
Object Type:
Report
System:
The UNT Digital Library
Experimental and Analytical Reactivity Studies of Clean Critical Stainless Steel Cores
ABS>The results are presented of critical water height measurements made on close-packed lattices of Spert III, highly enriched, plate-type, stainless- steel-clad fuel elements. Experiments were conducted with cores containing no control rods and with cores containing a single, fully-inserted control rod. The "clean critical" data obtained in these experiments were used to test the validity of various aspects of a four-group, diffusion theory analysis of the full scale Spert III reactor. The results of the analyses of the rod-free and single-rodded critical lattices show that for such stainless steel cores k/sub eff/ can be calculated to within 1% DELTA k and that the Spert III control rod worth is calculable to a few tenths % DELTA k. (auth)
Date:
June 16, 1961
Creator:
Spano, A. H.
Object Type:
Report
System:
The UNT Digital Library
SNAP 2 PRIMARY SYSTEM TEST-OBJECTIVES, SYSTEM DESCRIPTION, AND PROCEDURES
The SNAP-2 Primary System Test loop fabrication was completed with associated flight prototype components including reactor core and boiler mockups for volume and DELTA P simulation, CRU-IIII NaK pump, compact heater, and expansion compensator. A mobile loading system was designed and fabricated with the capability of cleaning the NaK prior to final loop sealing. Loop descriptions, test objectives, and operating procedures are presented. (auth)
Date:
June 12, 1961
Creator:
Kikin, G.M.
Object Type:
Report
System:
The UNT Digital Library
A Small Scale Countercurrent Liquid-Liquid Extractor
Details of design and operation are given for a laboratorysize, 20 - stage, multiple-contact, "unlimited-feed'', countercurrent-flow, liquid-liquid extractor. The apparatus consists mainly of glass parts that are joined by polyethylene tubing and are mounted in a steel cradle that can rotate on its horizontal axis. The reservoirs for feed liquids are an integral part of the assembly, and proper rotation of the assembly causes the flow of liquids to, through and from the extractor. Testing and developing of, and small scale production by, extraction systems are conveniently carried out in this extractor. Stagewise and product samples can be readily obtained for study of the extraction behavior of the components of a liquid-liquid system. (auth)
Date:
June 1, 1961
Creator:
Wilhelm, H. A.
Object Type:
Report
System:
The UNT Digital Library
Dynamic Simulation of Multi-Pass Pressurized Water Nuclear Power Plants by Analog Computer Techniques
A kinetic model of the primary loop of a multi-pass pressurized water reactor power plant is developed to evaluate, by analog computer techniques, the transient response characteristics under conditions of steam generator load and reactor control rod perturbations. Using the 2-pass 28 Mw(t) SM-2 reactor as a typical plant, transient behavior patterns are illustrated and examined for a variety of load inputs, variations in plant constants, and analog model simplifications. (auth)
Date:
June 1, 1961
Creator:
Brondel, J. O.
Object Type:
Thesis or Dissertation
System:
The UNT Digital Library
PROGRAM ODD--A ONE-DIMENSIONAL MULTIGROUP CODE FOR THE IBM-7090 (ANP PROGRAM NO. 657)
The physical and mathematical reactor models which are used in Program ODD are discussed. In addition, the FORTRAN II source program listings, decimal data input sheets, and input and output for a sample case are given. Program ODD was designed to raake use of the Revised Nuclear Data System at ANPD which consists of twenty-five energy group cross-section data including high energy inelastic scattering matrices, resonance parameters for the resolved resonances, and thermalization scattering matrices for the near thermal energy region. The most unique aspect of the program is the mathematical technique employed for eliminating inner iterations and slow convergenc rates occasioned by the up- scattering'' in the thermalization region of the energy lattice. Direct inversion of the energy matrix coupling the thermal and last four epitherma groups provides simultaneous consistent solutions for thes groups within each power iteration. (auth)
Date:
June 30, 1961
Creator:
Fischer, P.G.; Wenstrup, F.D. & Hoffman, T.A.
Object Type:
Report
System:
The UNT Digital Library
A Neutron Diffraction Study of Krypton in the Liquid State
A study was made of the neutron diffraction patterns obtained from Kr liquid under seventeen conditions of temperature and pressure at 117 to 210 deg K. The low temperatures were used because the diffraction patterns and the corresponding radial distribution functions are more sharply defined near the liqdid triple point. (J.R.D.)
Date:
June 1, 1961
Creator:
Clayton, G.T. & Heaton, L.
Object Type:
Report
System:
The UNT Digital Library
Preliminary Results of High-Temperature Bare U$sup 235$-C Critical Assembly Measurements
The influence of temperature on the critical buckling or bare graphite assemblies with various carbon-to-uranium235 molar ratios has been measured. A range from l185: 1 to 2l,690: 1 was covered, for 45 to 1205'F. Preliminary results indicate that the fractional rate of change of critical buckling with core temperature varies monotonically with C/U2as ratio by a factor of five over the factor-of-eighteen range in gross C/U2as ratio. This quantity appears to approach asymptotically a value near 2%/l00"F at very high C/U2ss ratios. (auth)
Date:
June 1, 1961
Creator:
Finke, R. G.
Object Type:
Report
System:
The UNT Digital Library
Force Multiplier for Use With Master Slaves
A force multiplier was designed. This piece of equipment was made to increase the gripping force presently available in the Model 8 master slave. The force multiplier described incorporates a clamp which can be quickly attached to and detached from the master slave hand. (auth)
Date:
June 1, 1961
Creator:
Miles, L. E.; Parsons, T. C. & Howe, P. W.
Object Type:
Report
System:
The UNT Digital Library
PL FINAL DESIGN REPORT. VOLUME V. CORE DRAWINGS AND SPECIFICATIONS
Drawings andd specifications for the PL reactor core are given. The requirements for the material procurement, fabrication, testing, and inspection of one P L-2 reactor core and spares are listed. (M.C.G.)
Date:
June 30, 1961
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Pressure Drop of Multirod Elements With Helical Spring Spacers
The pressure drop of a new fuel element design concept of spacing rods by means of helical wire springs was investigated experimentally and analytically. Extensive single- and two-phase pressure drop data at 1,000 psia were obtained for one flow geometry and helical spring spacer. Test conditions ranged from 0.7 to 1.2 x 10/sup 6/ lb/hr ft/sup 2/ in mass velocity and from 0 to 15% in quality. The effect of the specific spring which was tested was to increase the over-all pressure drop by 70%. A general analytical model was developed to predict the pressure drop of an element with helical spring spacers when the pressure drop without springs is known. The accuracy of the model, compared to the experimental data, was better than plus or minus 22%. The analytical model allows determination and evaluation of an optimum helical spring spacer design, so that pressure drop will not be a serious disadvantage. (auth)
Date:
June 1, 1961
Creator:
Quinn, E. P.
Object Type:
Report
System:
The UNT Digital Library
SL-1 ANNUAL OPERATING REPORT, FEBRUARY 1960-JANUARY 3, 1961
>The plant was operated from Feb. l959 until Jan. 3, 1961 when a nuclear excursion rendered the plant inoperable, A summary of operations is presented for the period Feb. 1960 until the incident. Plant operational tinwe during the period was 78% of that available. Satisfactory operation of conventional plant equipment was observed during 196 startups and scrams. An inspection of fuel assemblies after 700 Mwd showed no damage except to the B-Al strips attached to these assemblies. A greater than predieted gain in core reactivity was attributed to boron loss. Control rod sticking occurred 53 times ln 2,730 cases of operation. Test results on a PL-type condenser showed that its performance meets design rating. Steam quality remained greater than 99% for power levels up to 4.7 Mw(t) and radiation in the main steam system continued to be satisfactorily low. Waste handling requirements were increased but posed no problem. (J.R.D.)
Date:
June 15, 1961
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Mortality in Small Animals Exposed in a Shock Tube To "Sharp"-Rising Overpressures of 3-4 Msec Duration. Technical Progress Report
A total of 661 animals was exposed to sharp''-rising overpressures of 3 to 4 msec duration using a shock tube of novel design which produced a pressure pulse similar to that obtained with high explosives. The reflected shock overpressures associated with 50% lethality were 29.0, rabbit, respectively. Other observations included the time of death in mortally wounded animals and gross pathological lesions likely to contribute to mortality. Selected data from the literature bearing upon the influence of overpressure and pulse duration on lethality were reviewed. These included pulse durations ranging from less than 1 msec to 8 sec. The critical pulse duration, that duration shorter than which the overpressures required for mortality increases sharply, was noted to depend upon animal size and to be of the order of many hundreds of microseconds to very few milliseconds for smaller'' animals and a few to many tens of milliseconds for larger'' animals. (auth)
Date:
June 15, 1961
Creator:
Richmond, D. R.; Goldizen, V. C.; Clare, V. R.; Pratt, D. R.; Sherping, F.; Sanchez, R. T. et al.
Object Type:
Report
System:
The UNT Digital Library
TWENTY-FIVE GROUP REACTOR NUCLEAR DATA TAPE NEUTRON CROSS SECTIONS
A compilation is presented, in the twenty-five group Reactor Nuclear Data Tape format, of neutron cross sections for elements of major interest for GE- ANPD reactor analysis. The tabulated data are a reproduction of neutron cross section information contained on the Reactor Nuclear Data Tape, which was recently prepared. A brief outline of methods used in processing of the cross sections is also included. (auth)
Date:
June 13, 1961
Creator:
Zwick, J. W. & Kostigen, T. J.
Object Type:
Report
System:
The UNT Digital Library
Flux and Power Distributions for the SM-2 Reference and Critical Experiment Cores
A detailed analysis was made of the power distributions in the SM-2 experimental core at 68 deg F and the SM-2 reference core at 510 deg F. This analysis supersedes the power distribution calculations presented in APAE No. 69. The calculated distributions for the experimental core were normalized to measured data wherever possible in order to obtain corrections factors for application to the reference core. The over-all power distributions were calculated by synthesis of one-dimensional axial calculations with twodimensional radial calculations. The variation of the power distribution with fuel burnup is also presented. In order to improve the agreement between measured and calculated axial power distributions, flux-weighting the nuclear parameters in the rods-in and rods-out regions was investigated. (auth)
Date:
June 30, 1961
Creator:
Fried, B. E.; Alford, M. R. & Oggerino, J. P.
Object Type:
Report
System:
The UNT Digital Library
PL FINAL DESIGN REPORT. VOLUME III. PLANT EQUIPMENT SPECIFICATIONS
Specifications for the plant equipment for a P L-2 nuclear power plant are given. (M.C.G.)
Date:
June 30, 1961
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
[Man in Kitchen with Cup]
Photograph of a man in kitchen placing a cup back inside of a cupboard. The man is dressed in a dress shirt and slacks.
Date:
June 1961
Creator:
unknown
Object Type:
Photograph
System:
The UNT Digital Library
[Family Sharing a Meal]
Photograph of a family sharing a meal in a graduate apartment kitchen. There is two small boys, a man, and a woman around the table.
Date:
June 1961
Creator:
unknown
Object Type:
Photograph
System:
The UNT Digital Library
[Industrial Arts Building]
Photograph of the exterior of the Industrial Arts Building.
Date:
June 1961
Creator:
unknown
Object Type:
Photograph
System:
The UNT Digital Library