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SIMMER as a safety analysis tool

Description: SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program.
Date: January 1, 1982
Creator: Smith, L. L.; Bell, C. R.; Bohl, W. R.; Bott, T. F.; Dearing, J. F. & Luck, L. B.
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Partner: UNT Libraries Government Documents Department
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Relevance of biotic pathways to the long-term regulation of nuclear waste disposal. A report on Tasks 1 and 2 of Phase I. [Shallow land burial]

Description: The purpose of the work reported here was to evaluate the relevance of biotic transport to the assessment of impacts and licensing of low-level waste disposal sites. Available computer models and their recent applications at low-level waste disposal sites are considered. Biotic transport mechanisms and processes for both terrestrial and aquatic systems are presented with examples from existing waste disposal sites. Following a proposed system for ranking radionuclides by their potential for bio… more
Date: July 1, 1982
Creator: McKenzie, D. H.; Cadwell, L. L.; Cushing, C. E. Jr.; Harty, R.; Kennedy, W. E. Jr.; Simmons, M. A. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Description: Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more
Date: January 1, 1982
Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.
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Partner: UNT Libraries Government Documents Department
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Effect of reactor coolant pumps following a small break in a pressurized water reactor

Description: Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the … more
Date: January 1, 1982
Creator: Elliott, J.L.; Lime, J.F. & Willcutt, G.J.E. Jr.
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Partner: UNT Libraries Government Documents Department
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Mathematical models and illustrative results for the RINGBEARER II monopole/dipole beam-propagation code

Description: RINGBEARER II is a linearized monopole/dipole particle simulation code for studying intense relativistic electron beam propagation in gas. In this report the mathematical models utilized for beam particle dynamics and pinch field computation are delineated. Difficulties encountered in code operations and some remedies are discussed. Sample output is presented detailing the diagnostics and the methods of display and analysis utilized.
Date: May 24, 1982
Creator: Chambers, F.W.; Masamitsu, J.A. & Lee, E.P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Validation of SSC using the FFTF natural-circulation tests

Description: As part of the Super System Code (SSC) validation program, the 100% power FFTF natural circulation test has been simulated using SSC. A detailed 19 channel, 2 loop model was used in SSC. Comparisons showed SSC calculations to be in good agreement with the Fast Flux Test Facility (FFTF), test data. Simulation of the test was obtained in real time.
Date: January 1, 1982
Creator: Horak, W.C.; Guppy, J.G. & Kennett, R.J.
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Partner: UNT Libraries Government Documents Department
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Preliminary flashing multiphase flow analysis with application to letdown valves in coal-conversion processes

Description: As part of the Oak Ridge National Laboratory's technical support to large coal liquefaction projects, attempts have been made to (1) develop the methodology for characterizing and predicting multicomponent, multiphase, non-Newtonian flow behavior within letdown valves and devices, and (2) analyze the fluid flow in the entire letdown region of the process. An engineering model that can be used in the analysis of multicomponent, multiphase, flashing, flowing systems has been developed. A pre… more
Date: September 1, 1982
Creator: Ott, L. J. & Khan, A. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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LMFBR system-wide transient analysis: the state of the art and US validation needs

Description: This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.
Date: January 1, 1982
Creator: Khatib-Rahbar, M.; Guppy, J.G. & Cerbone, R.J.
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Partner: UNT Libraries Government Documents Department
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Flow regimes

Description: Internal boundaries in multiphase flow greatly complicate fluid-dynamic and heat-transfer descriptions. Different flow regimes or topological configurations can have radically dissimilar interfacial and wall mass, momentum, and energy exchanges. To model the flow dynamics properly requires estimates of these rates. In this paper the common flow regimes for gas-liquid systems are defined and the techniques used to estimate the extent of a particular regime are described. Also, the current comput… more
Date: January 1, 1982
Creator: Liles, D.R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Sensitivity study of neutron transport through standard and rebar concrete

Description: An investigation is under way at ORNL to (1) develop a data base pertinent to the transport of neutrons through thick concrete shields, (2) use the data base in an energy group boundary selection and collapsing scheme, and (3) develop a simple methodology to access the data base to provide rapid solutions to practical shielding problems. This paper describes work carried out to fulfill objective (1), the work consisting of calculations of the transport of fission neutrons through 1- and 2-m-thi… more
Date: January 1, 1982
Creator: Bhuiyan, S. I.; Roussin, R. W. & Lucius, J. L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Diagnostics for multiple regression problems

Description: In the last 10 to 15 years there has been much work done in trying to improve linear regression results. Individuals have analyzed the susceptibility of least-squares results to values far removed from the center of the independent variable observations. They have studied the problem of heavy-tailed residuals, and they have studied the problem of collinearity. From these studies have come ridge regression techniques, robust regression techniques, regression on principal components, etc. However… more
Date: March 1, 1982
Creator: Daly, J.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effect of pressure on the transient swelling rate of oxide fuel. [LMFBR]

Description: An analysis of the transient swelling rate of oxide fuel, based on fission-gas bubble conditions calculated with the FRAS3 code, has been developed and implemented in the code. The need for this capability arises in the coupling of the FRAS3 fission-gas analysis code to the FPIN fuel-pin mechanics code. An efficient means of closely coupling the calculations of swelling strains and stresses between the modules is required. The present analysis provides parameters that allow the FPIN calculation… more
Date: April 1, 1982
Creator: Gruber, E.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Verification of the three-dimensional thermal-hydraulic models of the TRAC accident-analysis code. [PWR]

Description: The Transient Reactor Analysis Code (TRAC) being developed at Los Alamos National Laboratory provides a best-estimate prediction of the response of light water reactors or test facilities to postulated accident sequences. One of the features of the code is the ability to analyze the vessel and its heated core in three dimensions. The code is being used to analyze the results of tests in a large-scale reflood test facility built in Japan, known as the Cylindrical Core Test Facility (CCTF). Two t… more
Date: January 1, 1982
Creator: Motley, F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Seismic Safety Margins Research Program: Phase II program plan (FY 83-FY 84)

Description: The Seismic Safety Margins Research Program (SSMRP) is an NRC-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its goal is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis and design process and explicitly incl… more
Date: August 2, 1982
Creator: Bohn, M. P.; Bernreuter, D. L.; Cover, L. E.; Johnson, J. J.; Shieh, L. C.; Shukla, S. N. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Impact of source terms on distances to which reactor accident consequences occur

Description: Estimates of the distances over which reactor accident consequences might occur are important for development of siting criteria and for emergency response planning. This paper summarizes the results of a series of CRAC2 calculations performed to estimate these distances. Because of the current controversy concerning the magnitude of source terms for severe accidents, the impact of source term reductions upon distance estimates is also examined.
Date: January 1, 1982
Creator: Ostmeyer, R.M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)

Description: The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, an… more
Date: January 1, 1982
Creator: Williams, M.L.; Wright, R.Q.; Barhen, J. & Rothenstein, W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Implementation and application of RCP system at BNL

Description: The implementation of the RCPO1 Monte Carlo system at the Brookhaven National Laboratory Central Scientific Computing Facility is discussed. Initial benchmarking experience and an LWR core physics application will be described.
Date: January 1, 1982
Creator: Eisenhart, L.D. & Todosow, M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Compare containment subcompartment analysis code evaluation. [PWR; BWR]

Description: Nuclear power plant subcompartment analyses are required to determine the containment pressure distribution that might result from a loss-of-coolant accident. The pressure distribution is used to calculate structural and mechanical design loads. The COMPARE code is used widely to perform subcompartment analysis. However, several simplifying assumptions are utilized to facilitate solution of the complex transient, two-phase, multidimensional flow problem. In particular, it is assumed that the fl… more
Date: January 1, 1982
Creator: Gido, R.G. & Koestel, A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Implementation and application of RCP system at BNL

Description: The implementation of the RCP01 Monte Carlo system at the Brookhaven National Laboratory Central Scientific Computing Facility is described. Newly written system independent software is stressed. Finally, initial benchmarking experience and an LWR core physics application are described.
Date: January 1, 1982
Creator: Eisenhart, L.D. & Todosow, M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Computational tools for pulverized-coal combustion. Third quarterly report, October 1981-December 1981

Description: This report summarizes the work conducted to develop and verify a computer code capable of modeling the major aspects of pulverized-coal combustion. Achieving this objective will lead to design methods applicable to industrial and utility furnaces. The combustion model (COMO) consists of a number of relatively independent modules that represent the major processes involved in pulverized-coal combustion. These modules will be continually upgraded over the course of the contract. Three distinct v… more
Date: January 1, 1982
Creator: Oberjohn, W. J.; Cornelius, D. K.; Fiveland, W. A.; Schnipke, R. J. & Wang, J. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Validation of ANS-5. 1 as the decay heat standard at the Savannah River Plant

Description: The Savannah River Laboratory (SRL) is upgrading the methods used to predict the post shutdown decay heat of the Savannah River reactors by implementing procedures based on the ANS Decay Heat Power in Light Water Reactors standard. This approach takes advantage of the large volume of research used in developing the standard and establishes compatibility with the nuclear industry. To qualify the decay heat standard for use, a series of comparisons were made between detailed decay heat calculatio… more
Date: January 1, 1982
Creator: Apperson, Jr, C. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Reactivity analysis of core distortion effects in the FFTF

Description: An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analys… more
Date: January 1, 1982
Creator: Knutson, B. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Use of separate-effects experiments in verification of system thermal-hydraulics

Description: In recent years, a number of advanced, best-estimate systems codes such as TRAC and RELAP5 have been developed in order to accurately predict the consequences of various postulated accidents and transients in Light Water Reactor (LWR) systems. Although these codes had to go through some verification or assessment during the developmental stage, it has been recognized that an independent assessment of these codes is necessary before they should be applied to any decision making process. The USNR… more
Date: January 1, 1982
Creator: Saha, P.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Curium-245 and curium-247 neutron cross sections between 10 keV and 10 MeV

Description: The optical model code 2PLUS and the statistical model codes COMNUC and CASCADE were used to compute neutron cross sections for Cm-245 and Cm-247 between 10 keV and 10 MeV. Cross sections for elastic and inelastic scattering, radiative capture, fission, and the (n,2n) reactions were computed. The parameters for the fission model were selected to yield agreement with the cross sections from the Physics-8 bomb shot. Pu-239 cross sections were calculated and compared with existing cross section ev… more
Date: July 1, 1982
Creator: Clifford, L.R. & McCrosson, F.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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