541 Matching Results

Search Results

Advanced search parameters have been applied.
in subjects: "21 Specific Nuclear Reactors And Associated Plants"  
Results: 1 - 24 next
open access

NRC plan for cleanup operations at Three Mile Island Unit 2

Description: This NRC Plan, which defines NRC's functional role in cleanup operations at Three Mile Island Unit 2 and outlines NRC's regulatory responsibilities in fulfilling this role, is the first revision to the initial plan issued in July 1980 (NUREG-0698). Since 1980, a number of policy developments have occurred which will have an impact on the course of cleanup operations. This revision reflects these developments in the area of NRC's review and approval process with regard to cleanup … more
Date: February 1, 1982
Creator: Lo, R. & Snyder, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Safety rod latch inspection

Description: During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and met… more
Date: February 1, 1992
Creator: Leader, D.R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

DOE role in nuclear policies and programs: official transcript of public briefing. Addendum December 13, 1977, Washington, D. C

Description: A total of 24 questions were read into the official record at the public briefing on nuclear policies and programs. The answers published were researched and written by personnel of DOE's Office of Energy Research, Office of Energy Technology, and the Secretary's Office. A few questions were sent to the Nuclear Regulatory Commission for review and for preparation of answers.
Date: February 1, 1978
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR]

Description: The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-… more
Date: February 1, 1978
Creator: Benedetti, R. L.; Lords, L. V. & Kiser, D. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Structural materials for breeder reactor cores and coolant circuits

Description: The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these compo… more
Date: February 1, 1984
Creator: Diercks, D. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

Description: The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium … more
Date: February 1, 1978
Creator: Jensen, D. D.; Haire, M. J. & Ballagny, A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Quarterly Progress Report on Fission Product Behavior in LWRs for the Period October-December 1977

Description: Analysis of release data obtained during High Burnup Fuel Test 10 (HBU-10) has been completed. In this test the fuel rod segment was ruptured by internal pressurization at 900/sup 0/C, at which time the temperature was rapidly increased to 1200/sup 0/C and maintained at this temperature for 10 min. Approximately 0.061% of the total cesium inventory in the rod segment was released; this was accompanied by the release of about 1.69% of the total /sup 85/Kr inventory. Moreover, about 0.022% of the… more
Date: February 1, 1978
Creator: Malinauskas, A. P.; Lorenz, R. A.; Collins, J. L.; Osborne, M. F.; Whatley, S. K. & Towns, R. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Silicon mass transfer in sodium loops and the resulting/thermal hydraulic effects. [LMFBR]

Description: The element silicon in the surface of new, 300 series stainless steel has been shown to rapidly dissolve in sodium above 525/sup 0/C. It deposits in slightly cooler regions as a crystalline compound with sodium and oxygen. In tests, the deposits have caused increases in hydraulic friction factor (hence, increased pressure loss) of up to 300% at Reynolds Numbers of 14/sup 4/ to 10/sup 5/.Also, they have contributed to local losses of heat transfer rate to 1/10 the original value, at a Reynolds N… more
Date: February 1, 1980
Creator: Yunker, W.H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

Description: The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydr… more
Date: February 1, 1984
Creator: Liles, D.R. & Mahaffy, J.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of LOFT steam generator main feed piping loop seal modification

Description: The stress analysis is presented for the proposed loop seal modification to the LOFT Steam Generator Main Feed Piping. THE SAP IV finite element computer program was used to analyze normal, upset, emergency, and faulted conditions. Results of the analysis indicate that the modified main feed piping system will satisfy all structural adequacy criteria specified in Subarticle NC-3650 of the ASME Boiler and Pressure Vessel Code. Results also show that the isolation snubber configuration, specified… more
Date: February 14, 1978
Creator: Nitzel, M.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

LOFT experimental measurements uncertainties analyses. Volume XVI. LOFT three-beam gamma densitometer system

Description: The magnitudes of various uncertainties in the loss-of-fluid test gamma densitometer measurements have been estimated. The dominant error in the estimate of the density profile and the average density is the extrapolation from the three chordal average density values to the total density profile. The primary uncertainty in each chordal average density measurement is the random noise inherent in the radiation process.
Date: February 1, 1978
Creator: Lassahn, G. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Sodium-water reaction acoustic noise for liquid phase injections. [LMFBR]

Description: Data on liquid and steam injections into sodium were recorded during a series of wastage experiments. These data are analyzed for acoustic power and spectral characteristics, expanding the data base up to 10 gm/sec injection rates from the earlier 0.5 gms/sec. No significant difference in acoustic power was measured between low temperature steam and liquid injections for the same mass flowrates. The bandwidth for steam injections is broader than for liquid injections. Reaction product depositio… more
Date: February 1, 1981
Creator: Callis, K. R.; Greene, D. A. & Malovrh, J. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Fast Flux Test Facility loose-parts monitor

Description: This paper summarizes the development testing in progress at the FFTF to determine the effectiveness of high temperature microphones as acoustic monitors in the upper plenum of the FFTF. The specific goal of this testing is development of an automated loose parts monitor for the upper plenum. A description of the acoustic probe is included, as well as a discussion of the signal processing. A summary of the results to date is also given.
Date: February 1, 1982
Creator: Sloan, W.R.; King, D.C. & Robles, M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Measured residual stresses in overlay pipe weldments removed from service

Description: Surface and throughwall residual stresses were measured on an elbow-to-pipe weldment that had been removed from the Hatch-2 reactor about a year after the application of a weld overlay. The results were compared with experimental measurements on three mock-up weldments and with finite-element calculations. The comparison shows that there are significant differences in the form and magnitude of the residual stress distributions. However, even after more than a year of service, the residual stres… more
Date: February 1, 1985
Creator: Shack, W.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Savannah River Laboratory monthly report, February 1992

Description: This report is a progress report for the Savannah River Laboratory for the month of February 1992. The progress and activities in six categories were described in the report. The categories are reactor, tritium, separations, environmental, waste management, and general. Each category described numerous and varied activities. Some examples of these activities described are such things as radiation monitoring, maintenance, modifications, and remedial action.
Date: February 1, 1992
Creator: Ferrell, J.M. (comp.) & Ice, L.W. (ed.)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

BWR refill-reflood program: core spray distribution experimental task plan

Description: An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.
Date: February 1, 1981
Creator: Eckert, T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Method for determining the uncertainty of gap conductance deduced from measured fuel centerline temperatures. [BWR]

Description: The paper describes the method which was developed to determine the uncertainties of gap conductances deduced from measured fuel centerline temperatures of NRC-RSR/BPNL fuel rods irradiated in the Halden Boiling Water Reactor. The ..integral..k(t)dt method is used to calculate the fuel surface temperature from the measured fuel centerline temperature and the fuel thermal conductivity. The gap conductance is calculated from the fuel surface temperature, the calculated cladding inside surface tem… more
Date: February 1, 1977
Creator: Hann, C. R.; Lanning, D. D.; Marshall, R. K.; Olsen, A. R. & Williford, R. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

Description: This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, ac… more
Date: February 1, 1988
Creator: Weiss, A. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

Description: Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of… more
Date: February 14, 1978
Creator: Leon, D. M.; White, M. D.; Moore, P. A. & Hedrick, R. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

PWR Blowdown Heat Transfer Separated-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 102. [PWR]

Description: Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 102, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 102 was conducted to investigate the thermal-hydraulic response of bundle 1 under full-power st… more
Date: February 1, 1978
Creator: Clemons, V. D.; White, M. D.; Moore, P. A. & Hedrick, R. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR]

Description: Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, … more
Date: February 1, 1978
Creator: Esparza, V.; Collins, B.L.; Sackett, K.E. & Coppin, C.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

Description: Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilizati… more
Date: February 1980
Creator: McDonald, C. F.; Vrable, D. L.; Van Hagan, T. H.; King, J. H. & Spring, A. H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

LOFT emergency core coolant thermal analysis Class I review

Description: The LOFT Emergency Core Coolant Piping Thermal Analysis was checked to insure that the calculations made would conservatively satisfy the requirements in the LOFT technical specifications. Some of the boundary conditions used have not been shown to be conservative and require review and possible re-analysis. One of the thermal models used could not be clearly related to a specific part of the piping geometry and requires further explanation. The remainder of the models, the use of the SIMIR cod… more
Date: February 3, 1978
Creator: Kinnaman, T.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
Back to Top of Screen