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This report provides documentation of the necessary source code modifications made to the TRAC-PF1/MOD2 code version 5.4.28a developed at Los Alamos National Laboratory.

This report documents the approaches taken in establishing a 1-dimensional integrated blanket system model using the TRAC code, developed by Los Alamos National Laboratory.

The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization.

This report describes the expanded set of test problems that were created to augment the existing Transient Reactor Analysis Code (TRAC)-P standard matrix of test problems. The expanded test problems were created to support the TRAC-P modernization effort. In most cases, these test problems were modified or expanded versions of problems in the TRAC Standard Test Matrix. A problem description is included for each problem added to the Standard Test Matrix. In this description, the details regarding modifications of the original test problem are included, as well as the observed problem results. This expanded test set will be used to verify that the predicted results for the modernized version of TRAC-M/f90 are null relative to the archival version of TRAC-P/MOD2 (Ver. 5.4.25), the latter being the base version on which work on the modernized code began. The problems described in this document increase the percentage of executable source coding that is activated when the Standard Test Matrix is run from 71% to 83.5%. A brief review of the characteristics of the portions of the source code that are not activated when running the expanded Standard Test Set also is provided. With a few exceptions, the author has concluded that the cost to increase the fraction of coding activated by the Standard Test Set further is high relative to the increased coverage that can be attained. Therefore, the author recommends that this activity be concluded.

Update FXCFM corrects five areas of the TRAC-P choked-flow model that address TRAC-P Trouble Report items 235, 259, and 260. Knolls Atomic Power Laboratory, Marviken, Edwards, and Scientech critical-flow test problems were used to investigate the report errors and further errors that were found and to verify their correction.

This report summarizes the implementation of tools within Version 5.4.25R10 of the Transient Reactor Analysis Code (TRAC); this implementation allows the semiautomated optimization of the reflood constitutive package. The tools included a software package external to TRAC that used a line search method to minimize a generic function value given the function`s partial derivative vector with respect to a set of closure coefficients used within TRAC`s reflood model. Within TRAC, the generic function was a normalized penalty function dependent on time averaged calculated values of vapor temperature, vapor void fraction, wall to a fluid heat transfer rate (or wall temperature), and the respective steady state data. The penalty function was implemented only for a one dimensional vessel configuration because the available reflood data were taken primarily from postcritical heat flux tube experiments.

This document briefly describes the elements of the Nuclear Regulatory Commission`s (NRC`s) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC`s software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort.

The FILL component in TRAC-P defines a phasic-velocities boundary condition or total-mass-flow boundary condition. FILL option IFTY = 10 defines the total-mass flow and its composition for flow donoring from the FILL to its adjacent component by signal variables and/or control blocks. For flow from the adjacent component to the FILL component, the phasic densities of the adjacent-component cell need to be upstream donored by the IFTY = 10 option total-mass flow. Instead, the FILL-cell phasic densities are being downstream donored incorrectly by the IFTY = 10 option total-mass flow in determining the FILL-junction phasic velocities. Using the wrong donored phasic densities caused the phasic velocities determined from the total-mass flow to be evaluated incorrectly. Five errors related to phasic-density donoring into a FILL- or BREAK-component cell are corrected by update FXFILL. Seven versions of a new test problem test these corrections and show that the errors of trouble reports 189 and 190 no longer exist in TRAC-P.

New ``Sum N`` and ``Multiply N`` control blocks have been programmed by update SUMNMULN in TRAC-P Version 5.4.28. They define N signal variables and/or control blocks whose values are to be summed with optional weighting factors or multiplied, respectively.

The DSTEP = {minus}99 and TIMET {ge} 0.0 s feature in TRAC-P has been tested on the W4LOOP and W4LOOPR standard test problems and was found to give different W4LOOPR solutions for different values of TIMET when all these cases should have given the same solution. Update FXTIME has corrected part of the cause of that difference and reduced the difference by an order of magnitude. The error associated with the remaining difference has been documented in Trouble Report 238.

This is the final report of a six-month, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Nuclear-plant training simulators employ simplified one-dimensional thermal-hydraulics codes because of the demands to run in real time and with limited computing power. The objective of this project was to investigate the feasibility of using the advanced Transient-Reactor Analysis Code (TRAC) in a simulator to increase the fidelity of a simulator. Many issues need to be addressed to take such a complex code from a batch engineering environment to a real-time environment. Working with simulator vendor, GSE, the authors investigated the technical issues relating to integrating TRAC into a real-time environment. They also modified a nuclear power plant model for simulator purposes and investigated its performance in real time.

A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers.

The Transient Reactor Analysis Code (TRAC)-M system-wide and component data structures are to be reimplemented by using the new features of Fortran 90 (F90). There will be no changes to the conceptual design, data flow, or computational flow with respect to the current TRAC-P, except that readability, maintainability, and extensibility will be improved. However, the task described here is a basic step that does not meet all future needs of the code, especially regarding extensibility. TRAC-M will be fully functional and will produce null computational changes with respect to TRAC-P, Version 5.4.25; computational efficiency will not be degraded significantly. The existing component and functional modularity and possibilities for coarse-grained parallelism will be retained.

XTV is an X-Windows based Graphical User Interface for viewing results of Transient Reactor Analysis Code (TRAC) calculations. It provides static and animated color mapped visualizations of both thermal-hydraulic and heat conduction components in a TRAC model of a nuclear power plant, as well as both on-screen and hard copy two-dimensional plot capabilities. XTV is the successor to TRAP, the former TRAC postprocessor using the proprietary DISSPLA graphics library. This manual describes Version 2.0, which requires TRAC version 5.4.20 or later for full visualization capabilities.

The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA.

Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in TRAC-P is also given.

This report describes 15 different test problems in the TRAC-P (Transient Reactor Analysis Code) standard test matrix of 42 test-problem calculations. Their TRACIN input-data files are listed in Appendix A. The description of each test problem includes the nature of what the test problem models and evaluates, the principal models of TRAC-P that the test problem serves to verify or validate, and the TRAC-P features and options that are being involved in its calculation. The test-problem calculations will determine the effect that changes made to a TRAC-P version have on the results. This will help the developers assess the acceptance of those changes to TRAC-P.

A passive system can provide engineered safety features enhancing safety system reliability and plant simplicity. Toshiba has conducted the test Program to demonstrate the feasibility of the SBWR passive safety system using a full-height, integral system test facility GIRAFFE. The test facility GIRAFFE models the SBWR in full height to correctly present the gravity driving head forces with a 1/400 volume scale. The GIRAFFE test Program includes the certification tests of the passive containment cooling system (PCCS) to remove the post-accident decay heat and the gravity driven cooling system (GDCS) to replenish the reactor coolant inventory during a LOCA. The test results have confirmed the PCCS and GDCS design and in addition, have demonstrated the operation of the pCCS with the presence of a lighter-than-steam noncondensable as well as with the presence of a heavier-than-steam, noncondensable. The GIRAFFE test Program has also provided the database to qualify a best estimate thermal-hydraulic computer code TRAC. The post test analysis results have shown that TRAC can accurately predict the PCCS heat removal Performance and the containment pressure response to a LOCA. This paper summarizes the GIRAFFE test results to investigate post-LOCA PCCS heat removal performance and post-test analysis using TRAC.

An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their {und W}COBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse.

The sample data for selection of 11 radionuclides and 24 chemical analytes were extracted from six separate sample data sets, were arranged in a tabular format and were plotted on scatter plots for all of the 149 single-shell tanks, the 24 double-shell tanks and the four aging waste tanks. The solid and liquid sample data was placed in separate tables and plots. The sample data and plots were compiled from the following data sets: characterization raw sample data, recent core samples, D. Braun data base, Wastren (Van Vleet) data base, TRAC and HTCE inventories. This document is Volume I of the Letter Report entitled Tank Waste Source Term Inventory Validation.

Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes.

In March of 1988 the LaSalle 2 BWR, while at about 85 percent power, was exposed to a loss of both recirculation pumps providing drive flow to the jet pumps. Within a few minutes the reactor power began to oscillate, resulting in an overpower scram. This report presents results of calculations performed with the TRAC-BF1 code to assess the capability of the code to calculate the observed behavior of the LaSalle plant during the event.

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector Lmore » Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification.« less

This document is a set of detailed instructions and guidelines to aid users in constructing and interpreting FLOWTRAN-TF input and output files for version 1.2 of the source code. The document assumes the user is familiar with the FLOWTRAN-TF Software Design report, SRS fuel assembly hardware, and two-phase flow. General code capabilities and input options are summarized. Then, detailed instructions for creating and interpreting code input files are given next. A sample input deck and corresponding output files are listed for reference and illustration. FLOWTRAN-TF is a two-phase thermal-hydraulics code of similar technology to existing commercial reactor codes such as RELAP and TRAC but customized for Savannah River Site applications. The code may be used to simulate solid components, fluid coolant flow and solid-fluid heat transfer, or fluid flow only (adiabatic flow channels). Pure component water or two-component air-water flows may be modeled. A variety of materials may be chosen for the solid tubes separating flow channels. FLOWTRAN-TF is fundamentally a transient analysis tool.

FLOWTRAN-TF was created to analyze an individual Mk22 fuel assembly during a large break Loss Of Coolant Accident (LOCA) scenario involving the Savannah River Site K-reactor after the initial few seconds of the transient. During the initial few seconds reactor cooling is limited by the static or Ledinegg flow instability phenomenon. The predecessor FLOWTRAN code was developed to analyze this portion of a LOCA. In the several seconds following the break, a significant fraction of the reactor coolant inventory leaks out the break, Emergency Cooling System (ECS) flow is initiated, and air enters the primary coolant circulation loops. Reactor fuel assemblies are cooled by a low flowrate air-water downflow. Existing commercial nuclear industry thermal-hydraulic codes were judged inadequate for detailed modeling of a Mk22 fuel assembly because the application involves a ribbed annular geometry, low pressure, downflow and an air-water mixture. FLOWTRAN-TF is a two-phase thermal-hydraulics code of similar technology to existing commercial codes such as RELAP and TRAC but customized for Savannah River Site applications. The main features and capabilities of FLOWTRAN-TF are detailed Mk22 fuel assembly ribbed annular geometry; conjugate heat transfer; detailed neutronic power distribution; three-dimensional heat conduction in Mk22 fuel and target tubes; two-dimensional coolant flowmore » in channels (axial, azimuthal); single-phase and/or two-phase fluid (gas, liquid and/or gas-liquid); two-component (air, water); constitutive models applicable to low pressure air-water downflow in ribbed annular channels. The design of FLOWTRAN-TF is described in detail in this report which serves as the Software Design Report in accordance with Quality Assurance Procedure IV-4, Rev. 0 Software Design and Implementation'' in the 1Q34 manual.« less

A posttest calculation and analysis of Multi-Loop Integral System Test 320201, a small-break loss-of-coolant accident (SBLOCA) test with a scaled 50-cm[sup 2] cold-leg pump discharge leak, has been completed and is reported herein. It was one in a series of tests, with leak size varied parametrically. Scaled leak sizes included 5, 10, (the nominal, Test 3109AA), and 50 cm[sub 2]. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, interruption of loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vessel vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator (SG) secondaries was used after SG-secondary refill; and symmetric SG pressure control was also used. The sequence of events seen in this test was similar to the sequence of events for much of the nominal test except that events occurred in a shorter time frame as the system inventory was reduced and the system depressurized at a faster rate. The calculation was performed using TRAC-PFL/MOD 1. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. We believemore » that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

The ability to predict the behavior of reactor components to varying coolant flow scenarios constitutes a necessary skill for assessing reactor safety. One tool for performing these calculations is the Transient Reactor Analysis Code (TRAC). In order to benchmark the code, the Safety Analysis Group of SRL requested the Equipment Engineering Section (EES) of SRL to conduct a series of experiments to provide measurements of cooling parameters in a well defined physical system utilizing SRS reactor components. The configuration selected consisted of a short length of septifoil with both top and bottom fittings containing five simulated control rods in an {open_quotes}unseated{close_quotes} configuration. Varying power levels were to be supplied to the rods with 3.5 kilowatts per foot the value targeted for modelling during the computer runs. The septifoil segment was to be operated with no forced flow in order to evaluate thermal-hydraulic cooling. Parameters to be measured for comparison with code predictions were basic cooling phenomena, incidence of film boiling, water flow rate, pressure rise, and ratio of heat transfer through the wall of the assembly vs. heat transfer to axial water flow through the assembly. This report documents testing done with unimpeded flow into and out of the septifoilmore » in order to assess basic cooling phenomena, incidence of film boiling and pressure rise. Previous tests have evaluated water flow rate and the ratio of axial to azimuthal heat transfer.« less

The results produced by two of the WIND System atmospheric models, PUFF/PLUME and 2DPUF, were compared with a select group of eight MATS experiments to determine the performance of the models. Three of the MATS experiments employed TRAC vehicle sampling and the remaining five used a line of fixed samplers. The performance of the models was based on certain dispersion characteristics that are important in emergency response situations. Both PUFF/PLUME and 2DPUF were executed with the same source term and meteorological data. When the numerical results from the models were compared to the observed values from the MATS experiments, it was found that 2DPUF produced concentrations and plume widths that were closer to the observed values than PUFF/PLUME. Both models did not produce any bias in the values of the concentration when individual data points were examined; however, PUFF/PLUME consistently overpredicted the peak and total concentrations. 2DPUF did not exibit any bias in the peak and total concentrations. When wind direction errors were removed, 80--84% of the concentrations from PUFF/PLUME and 88% of the concentrations from 2DPUF where within a factor of 10 of the observed values. In some instances, both models were able to predict concentration values that weremore » comparible to a more complex, three-dimensional model called MATHEW-ADPIC. Considering all of the possible uncertainties associated with dispersion modeling, PUFF/PLUME and 2DPUF performed reasonably well. The differences between the dispersion forecasts made by PUFF/PLUME and 2DPUF and the observed surface tracer concentration are very similar to many other emergency response models based on the Gaussian assumption.« less

The TRAC-PF1/MOD1 code (TRAC) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). For this accident, TRAC is used to analyze only the first 5 seconds following the DEGB, which encompasses the Flow Instability (FI) phase of the DBA. The TRAC analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code. The quantification of uncertainty is an important element of determining safe operating power levels for SRS reactors. A detailed methodology for the determination of uncertainty for the FI phase of a DEGB LOCA has been developed. This report presents estimates of the uncertainty in the time-dependent plenum pressures for the DEGB LOCA calculated by TRAC. The plenum pressure uncertainty was estimated by means of comparing TRAC results with steady-state data measured in L Reactor, and confirmed by comparisons with transient LOCA results calculated by an independent group with the RELAP5 code. An overview of the limits methodology is given and discusses the L Reactor data. The methodology for estimating the plenummore » pressure uncertainty is presented along with the results.« less

An analysis of the OECD-LOFT-LP-SB-2 experiment making use of TRAC-PF1/MOD1 is described in the report. LP-SB2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3 inches equivalent diameter break in the hot leg of a commercial PWR operating at full power. The experiment was performed on 14 July 1983 in the LOFT facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Co-operation and Development (OECD). This analysis presents an evaluation of the code capability in reproducing the complex phenomena which determined the LP-SB-2 transient evolution. the analysis comprises the results obtained from two different runs. The first run is described in detail analysing the main variables over two time spans: short and longer term. Several conclusions are drawn and then a second run testing some of these conclusions is shown. All of the calculations were preformed at the United Kingdom Atomic Energy Establishment at Winfrith under the auspices of an agreement between the UKAEA (United Kingdom Atomic Energy Authority) and the Consejo de Seguridad Nuclear Espanol (CSN). 16 refs., 64 figs., 6 tabs.

A comparison of TRAC and RELAP5 predictions of steady-state and DEGB LOCA results (FI phase) for K-14.1 has been made. Both codes had been previously benchmarked against 1985 L Reactor AC Flow data and were under configuration control. The purpose of the code-to-code comparison is to provide insight on the transient uncertainty in TRAC plenum and tank bottom plenum pressures. The comparisons focus on LOCA results between 0.5 and 2.0 s, which is the primary period of interest for Flow Instability (FI) limits.

This report describes a general method for estimating the effect on the Reflood Phase PCT from systematic errors (biases) associated with the modelling of the ECCS and dissolved nitrogen, and the application of this method in estimating biases in the Reflood Phase PCT (second PCT) predicted by the TRAC/PF1/MOD1, Version 14.3. The bias in the second PCT due to the uncertainty in the existing code models for ECCS related phenomena is {minus}19{degree}K ({minus}34{degree}F). The negative bias implies that the code models for this phenomena are conservative. The bias in the second PCT due to the lack of modelling of dissolved N{sub 2} in the code is estimated to be 9.9{degree}K (17.8{degree}F). The positive bias implies that the absence of dissolved N{sub 2} model makes the code prediction of PCT non-conservative. The bias estimation in this report is a major exception among all other uncertainty and bias assessments performed in conjunction with the CSAU methodology demonstration, because this bias estimation benefitted from using full-scale test data from the full-scale Upper Plenum Test Facility (UPTF). Thus, the bias estimates presented here are unaffected by scale distortions in test facilities. Data from small size facilities were also available and an estimate of biasmore » based on these data will be conservative. 35 refs., 18 figs., 5 tabs.« less

The analysis of the Design Basis Loss of Coolant Acident (LOCA) for Savannah River Site (SRS) reactors involves the best estimate reactor system thermal-hydraulics code TRAC-PFI/MOD1. Power levels for the L-3.1 and P-10.2 subcycles were determined based, in part, on TRAC analyses of the first few seconds of a plenum inlet break LOCA. The TRAC code is currently being used to analyze reactor system response for the Double Ended Guillotine Break (DEGB) LOCA, the Expansion Joint Bellows Break LOCA, the Loss of Pumping Accident (LOPA), and the Pump Shaft Break event. Currently, the DEGB LOCA analysis is performed with TRAC only for the flow instability (FI) phase of the accident. This analysis provides input to the determination of operating power limits for the K-14.1 subcycle.

No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.

Analysis of the small, hot leg break, OECD LOFT Experiment LP-SB-1. using the best-estimate'' computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-1 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. 24 refs., 66 figs., 12 tabs.

Analysis of the small, cold leg break, OECD LOFT Experiment LP-SB-3 using the best-estimate computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-3 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. Results of the TRAC-PF1/MOD1 calculation are found to be generally consistent with those reported, by members of the OECD LOFT Program Review Group, in the LP-SB-3 Comparison Report.'' Overall trends with respect to pressure histories, minimum primary system mass inventory and accumulator behaviour are reasonably well reproduced by TRAC-PF1/MOD1. 17 refs., 26 figs., 3 tabs.

This volume provides production guidelines for the documentation of the Terrain-Responsive Atmospheric Code (TRAC). This documentation is divided into five categories: promotional, technical reference, system reference, bibliography, and user documentation. This format guide has been produced for TRAC editors and production personnel, as well as for the authors of TRAC documentation, so that they may have the insight as to how their finished product will look. This is done in the hope that by visualizing the finished, or near-finished product, authors will have at their disposal the format with which to conceptualize their raw material into published documentation. In creating five basic types of TRAC documents -- promotional, technical reference, system reference, bibliography and user documentation -- the editors have provided TRAC authors with the tools they need to channel their technical expertise into a larger, overall effort: the complete documentation of the Terrain-Responsive Atmospheric Code. 48 figs.

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes inmore » the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.« less

A summary of Transient Reactor Analysis Code Version PD2 (TRAC-PD2) calculations performed at the Idaho National Engineering Laboratory (INEL) is presented in this report as part of the US Nuclear Regulatory Commission's (NRCs) overall assessment program of TRAC-PD2. The calculated and measured parameters summarized in this report are break mass flow rate, primary coolant system pressure, reactor core flow rates, and fuel rod cladding temperatures. The data were obtained from seven tests that were performed at two test facilities. The tests were conducted to study the various aspects of cold leg break transients, including the effects of large and small beaks, and core reflood phenomena. User experience gained from the various calculations is also summarized. 42 figs., 10 tabs.

The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests.

TRAC is an advanced best-estimate computer code for analyzing postulated accidents in light water reactors. This paper gives a brief description of the code followed by comparisons of TRAC calculations with data from a variety of separate-effects, system-effects, and integral experiments. Based on these comparisons, the capabilities and limitations of the early versions of TRAC are evaluated.

This report documents the results of a 200%, double-ended, cold-leg-break, loss-of-coolant-accident (LOCA) calculation using the TRAC-PF1/MOD1 computer code. The reactor system represented a typical United States/Japanese pressurized water reactor with a 15 x 15 fuel bundle arrangement 12-ft long, four loops, and cold-leg Emergency Core Cooling (ECC) Systems. Conservation boundary and initial conditions were used. Reactor power was 102% of the 3250 MWt rated power, decay heat was set to 120% of American Nuclear Society Standard 5.1, highest core lifetime values for power peaking and fuel stored energy were used, and the LOCA occurred simultaneously with a loss of offsite power. Best estimate assumptions were used for the break flow model, fuel rod heat transfer and metal-water reaction correlations, and steady-state fuel temperature profiles. A flow blockage model, having the capability to account for the effects of cladding ballooning or rupturing, was not used. Except for these best estimate assumptions, the boundary and initial conditions were consistent with those used in licensing calculations. Maximum fuel rod temperatures were 1380 K (2020/sup 0/F) and 1040 K (1410/sup 0/F) on the hottest evaluation model rod and hottest best estimate rod, respectively. The high reported values or fuel cladding temperature were a directmore » consequence of the conservative boundary and initial conditions used for the calculation, primarily the 2% overpower condition, the core decay heat assumption, and the degraded ECCS. The calculation demonstrated successful core reflooding before 1478 K (2200/sup 0/F) cladding temperature was exceeded on any fuel rod. 7 refs., 47 figs., 5 tabs.« less

This report presents the results of the TRAC-P1A independent assessment analyses performed during calendar year 1979. These calculations were performed with the publicly released version of the code and include separate-effects tests for vessel level swell and large-scale critical flow, integral-effects tests for the blowdown/refill/reflood phases of the large-break LOCAs, and integral-effects tests for small-break LOCAs. Although the independent assessment analyses do not represent an exhaustive study of the full range of available facilities and tests, they do represent a rigorous test of the capabilities of the code. The results indicate that the code is directly applicable to LOCA analyses; several areas have been identified for improvement in future code development. 25 refs., 79 figs.

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling.

This report discusses preliminary studies of the Three Mile Island Unit 2 (TMI-2) accident based on available methods and data. The work reported includes: (1) a TRAC base case calculation out to 3 hours into the accident sequence; (2) TRAC parametric calculations, these are the same as the base case except for a single hypothetical change in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident; (3) fuel rod cladding failure, cladding oxidation due to zirconium metal-steam reactions, hydrogen release due to cladding oxidation, cladding ballooning, cladding embrittlement, and subsequent cladding breakup estimates based on TRAC calculated cladding temperatures and system pressures. Some conclusions of this work are: the TRAC base case accident calculation agrees very well with known system conditions to nearly 3 hours into the accident; the parametric calculations indicate that, loss-of-core cooling was most influenced by the throttling of High-Pressure Injection (HPI) flows, given the accident initiating events and the pressurizer electromagnetic-operated valve (EMOV) failing to close as designed; failure of nearly all the rods and gaseous fission product gas release from the failed rods is predicted to have occurred at about 2 hours and 30 minutes; cladding oxidation (zirconium-steam reaction) up to 3 hours resulted in the production of approximately 40 kilograms of hydrogen.

TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system. The constitutive correlations for shear and heat transfer in the boiling water reactor (BWR) version of TRAC are described. A new model, that accounts for the effect of phase and velocity profiles, has been developed for the interfacial shear and a new set of constitutive correlations are derived. Improvements have been made to the heat transfer in the area of subcooled boiling, boiling transition, and thermal radiation.