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Sample records for reactor system relap5

  1. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  2. A RELAP5/MOD3 simulation of loss of residual heat removal system after reactor

    SciTech Connect

    Tanrikut, A.; Heper, H.H.

    1995-12-31

    A computational investigation of the experiment concerning the loss of the residual heat removal system (RHRS) during reduced inventory operation was simulated using the RELAP5/ MOD3 thermal-hydraulic code. The experiment was conducted at the UMCP 2 x 4 integral test loop (University of Maryland) and consisted of two parts: loss of RHRS and loss of feedwater system. The objective of the work presented in this paper is to assess the capability of the RELAP5 code to capture the phenomena observed in the experiment during the boiler-condenser mode (BCM) and the loss of feedwater (LOFW) system transient.

  3. Modeling moving systems with RELAP5-3D

    DOE PAGESBeta

    Mesina, G. L.; Aumiller, David L.; Buschman, Francis X.; Kyle, Matt R.

    2015-12-04

    RELAP5-3D is typically used to model stationary, land-based reactors. However, it can also model reactors in other inertial and accelerating frames of reference. By changing the magnitude of the gravitational vector through user input, RELAP5-3D can model reactors on a space station or the moon. The field equations have also been modified to model reactors in a non-inertial frame, such as occur in land-based reactors during earthquakes or onboard spacecraft. Transient body forces affect fluid flow in thermal-fluid machinery aboard accelerating crafts during rotational and translational accelerations. It is useful to express the equations of fluid motion in the acceleratingmore » frame of reference attached to the moving craft. However, careful treatment of the rotational and translational kinematics is required to accurately capture the physics of the fluid motion. Correlations for flow at angles between horizontal and vertical are generated via interpolation where no experimental studies or data exist. The equations for three-dimensional fluid motion in a non-inertial frame of reference are developed. As a result, two different systems for describing rotational motion are presented, user input is discussed, and an example is given.« less

  4. Modeling moving systems with RELAP5-3D

    SciTech Connect

    Mesina, G. L.; Aumiller, David L.; Buschman, Francis X.; Kyle, Matt R.

    2015-12-04

    RELAP5-3D is typically used to model stationary, land-based reactors. However, it can also model reactors in other inertial and accelerating frames of reference. By changing the magnitude of the gravitational vector through user input, RELAP5-3D can model reactors on a space station or the moon. The field equations have also been modified to model reactors in a non-inertial frame, such as occur in land-based reactors during earthquakes or onboard spacecraft. Transient body forces affect fluid flow in thermal-fluid machinery aboard accelerating crafts during rotational and translational accelerations. It is useful to express the equations of fluid motion in the accelerating frame of reference attached to the moving craft. However, careful treatment of the rotational and translational kinematics is required to accurately capture the physics of the fluid motion. Correlations for flow at angles between horizontal and vertical are generated via interpolation where no experimental studies or data exist. The equations for three-dimensional fluid motion in a non-inertial frame of reference are developed. As a result, two different systems for describing rotational motion are presented, user input is discussed, and an example is given.

  5. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    SciTech Connect

    Griggs, D.P.

    1993-03-01

    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 L 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.

  6. System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

    SciTech Connect

    Meng Lin; Dong Hou; Zhihong Xu; Yanhua Yang; Ronghua Zhang

    2006-07-01

    Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, just can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is

  7. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  8. RELAP5 two-phase fluid model and numerical scheme for economic LWR system simulation

    SciTech Connect

    Ransom, V.H.; Wagner, R.J.; Trapp, J.A.

    1981-01-01

    The RELAP5 two-phase fluid model and the associated numerical scheme are summarized. The experience accrued in development of a fast running light water reactor system transient analysis code is reviewed and example of the code application are given.

  9. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    SciTech Connect

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes.

  10. RELAP5 MODEL OF THE DIVERTOR PRIMARY HEAT TRANSFER SYSTEM

    SciTech Connect

    Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

    2010-08-01

    This report describes the RELAP5 model that has been developed for the divertor primary heat transfer system (PHTS). The model is intended to be used to examine the transient performance of the divertor PHTS and evaluate control schemes necessary to maintain parameters within acceptable limits during transients. Some preliminary results are presented to show the maturity of the model and examine general divertor PHTS transient behavior. The model can be used as a starting point for developing transient modeling capability, including control system modeling, safety evaluations, etc., and is not intended to represent the final divertor PHTS design. Preliminary calculations using the models indicate that during normal pulsed operation, present pressurizer controls may not be sufficient to keep system pressures within their desired range. Additional divertor PHTS and control system design efforts may be required to ensure system pressure fluctuation during normal operation remains within specified limits.

  11. Extensions to SCDAP/RELAP5-3D for Analysis of Advanced Reactors

    SciTech Connect

    Harvego, Edwin Allan; Siefken, Larry James

    2003-04-01

    The SCDAP/RELAP5-3D code was extended to enable the code to perform transient analyses of advanced LWRs (Light Water Reactors) and HTGRs (High Temperature Gas Reactors). The extensions for LWRs included: (1) representation of micro-heterogeneous fuel varying in composition in the radial and axial directions, (2) modeling of two-dimensional radial/axial heat conduction for more accurate calculation of fuel and cladding temperatures during the reflood period of a large break loss-of-coolant accident (LOCA), (3) modeling of fuel-cladding interface pressure and fuel-cladding gap conductance, (4) representation of radial power profiles varying in a discontinuous manner in the axial direction, and (5) addition of material properties for fuel composed of mixtures of ThO2-UO2 and ThO2-PuO2. The extensions for HTGR analyses included: (1) modeling of the transient two-dimensional temperature behavior of graphite moderated reactor cores (pebble bed and block-type), reactor vessel, and reactor containment, (2) modeling of flow losses and convective heat transfer in pebble bed reactor cores, (3) modeling of oxidation of graphite components in reactor cores due to the ingress of air and/or water, and (4) modeling of the affect of oxidation on the composition of gases in the reactor system. The applications of the extended code to LWR analyses showed that advanced fuels intended for proliferation resistance and waste reduction could also be designed to produce calculated peak cladding temperatures during a large break LOCA less than the 1477 K acceptance criterion in 10 CFR 50.46. Fuels composed of ThO2-UO2 and ThO2-PuO2 are examples of such fuels. The applications of the extended code to HTGR analyses showed that: (1) HTGRs can be designed for passive removal of all decay heat, and (2)

  12. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  13. RELAP5 Model of the First Wall/Blanket Primary Heat Transfer System

    SciTech Connect

    Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

    2010-06-01

    ITER inductive power operation is modeled and simulated using a system level computer code to evaluate the behavior of the Primary Heat Transfer System (PHTS) and predict parameter operational ranges. The control algorithm strategy and derivation are summarized in this report as well. A major feature of ITER is pulsed operation. The plasma does not burn continuously, but the power is pulsed with large periods of zero power between pulses. This feature requires active temperature control to maintain a constant blanket inlet temperature and requires accommodation of coolant thermal expansion during the pulse. In view of the transient nature of the power (plasma) operation state a transient system thermal-hydraulics code was selected: RELAP5. The code has a well-documented history for nuclear reactor transient analyses, it has been benchmarked against numerous experiments, and a large user database of commonly accepted modeling practices exists. The process of heat deposition and transfer in the blanket modules is multi-dimensional and cannot be accurately captured by a one-dimensional code such as RELAP5. To resolve this, a separate CFD calculation of blanket thermal power evolution was performed using the 3-D SC/Tetra thermofluid code. A 1D-3D co-simulation more realistically models FW/blanket internal time-dependent thermal inertia while eliminating uncertainties in the time constant assumed in a 1-D system code. Blanket water outlet temperature and heat release histories for any given ITER pulse operation scenario are calculated. These results provide the basis for developing time dependent power forcing functions which are used as input in the RELAP5 calculations.

  14. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    SciTech Connect

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  15. Nuclear Hybrid Energy System Modeling: RELAP5 Dynamic Coupling Capabilities

    SciTech Connect

    Piyush Sabharwall; Nolan Anderson; Haihua Zhao; Shannon Bragg-Sitton; George Mesina

    2012-09-01

    The nuclear hybrid energy systems (NHES) research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation.

  16. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  17. Evaluation of the Use of Existing RELAP5-3D Models to Represent the Actinide Burner Test Reactor

    SciTech Connect

    C. B. Davis

    2007-02-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid that are not currently represented with internal code models, including axial and radial heat conduction in the fluid and subchannel mixing. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor. An evaluation was also performed to determine if the existing centrifugal pump model could be used to simulate the performance of electromagnetic pumps.

  18. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    SciTech Connect

    Marcum, W.R.; Woods, B.G.; Hartman, M.

    2008-07-15

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  19. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    SciTech Connect

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  20. Independent review of SCDAP/RELAP5 natural circulation calculations

    SciTech Connect

    Martinez, G.M.; Gross, R.J.; Martinez, M.J.; Rightley, G.S.

    1994-01-01

    A review and assessment of the uncertainties in the calculated response of reactor coolant system natural circulation using the SCDAP/RELAP5 computer code were completed. The SCDAP/RELAP5 calculation modeled a station blackout transient in the Surry nuclear power plant and concluded that primary system depressurization from natural circulation induced primary system failure is more likely than previously thought.

  1. RELAP5/MOD3 Analysis of Transient Steam-Generator Behavior During Turbine Trip Test of a Prototype Fast Breeder Reactor MONJU

    SciTech Connect

    Yoshihisa Shindo; Hiroshi Endo; Tomoko Ishizu; Kazuo Haga

    2006-07-01

    In order to develop a thermal-hydraulic model of the steam-generator (SG) to simulate transient phenomena in the sodium cooled fast breeder reactor (FBR) MONJU, Japan Nuclear Energy Safety Organization (JNES) verified the SG model using the RELAP5/MOD3 code against the results of the turbine trip test at a 40% power load of MONJU. The modeling by using RELAP5 was considered to explain the significant observed behaviors of the pressure and the temperature of the EV steam outlet, and the temperature of water supply distributing piping till 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results based on results of parameter study as the blow efficiency (release coef.) and heat transferred from the helical coil region to the down-comer (temperature heating down-comer tubes). It was found that the RELAP5/MOD3 code with a two-fluids model can predict well the physical situation: the gas-phase of steam generated by the decompression boiling moves upward in the down-comer tubes accompanied by the enthalpy increase of the water supply chambers; and that the pressure change of a 'shoulder' like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/MOD3 to SG modeling was confirmed by simulating the actual FBR system. (authors)

  2. Modeling and Analysis of a Lunar Space Reactor with the Computer Code RELAP5-3D/ATHENA

    SciTech Connect

    Carbajo, Juan J; Qualls, A L

    2008-01-01

    The transient analysis 3-dimensional (3-D) computer code RELAP5-3D/ATHENA has been employed to model and analyze a space reactor of 180 kW(thermal), 40 kW (net, electrical) with eight Stirling engines (SEs). Each SE will generate over 6 kWe; the excess power will be needed for the pumps and other power management devices. The reactor will be cooled by NaK (a eutectic mixture of sodium and potassium which is liquid at ambient temperature). This space reactor is intended to be deployed over the surface of the Moon or Mars. The reactor operating life will be 8 to 10 years. The RELAP5-3D/ATHENA code is being developed and maintained by Idaho National Laboratory. The code can employ a variety of coolants in addition to water, the original coolant employed with early versions of the code. The code can also use 3-D volumes and 3-D junctions, thus allowing for more realistic representation of complex geometries. A combination of 3-D and 1-D volumes is employed in this study. The space reactor model consists of a primary loop and two secondary loops connected by two heat exchangers (HXs). Each secondary loop provides heat to four SEs. The primary loop includes the nuclear reactor with the lower and upper plena, the core with 85 fuel pins, and two vertical heat exchangers (HX). The maximum coolant temperature of the primary loop is 900 K. The secondary loops also employ NaK as a coolant at a maximum temperature of 877 K. The SEs heads are at a temperature of 800 K and the cold sinks are at a temperature of ~400 K. Two radiators will be employed to remove heat from the SEs. The SE HXs surrounding the SE heads are of annular design and have been modeled using 3-D volumes. These 3-D models have been used to improve the HX design by optimizing the flows of coolant and maximizing the heat transferred to the SE heads. The transients analyzed include failure of one or more Stirling engines, trip of the reactor pump, and trips of the secondary loop pumps feeding the HXs of the

  3. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    Energy Science and Technology Software Center (ESTSC)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized watermore » reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.« less

  4. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    SciTech Connect

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized water reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.

  5. Assessment of RELAP5/MOD3.1 for gravity-driven injection experiment in the core makeup tank of the CARR Passive Reactor (CP-1300)

    SciTech Connect

    Lee, S.I.; No, H.C.; Bang, Y.S.; Kim, H.J.

    1996-10-01

    The objective of the present work is to improve the analysis capability of RELAP5/MOD3.1 on the direct contact condensation in the core makeup tank (CMT) of passive high-pressure injection system (PHPIS) in the CARR Passive Reactor (CP-1300). The gravity-driven injection experiment is conducted by using a small scale test facility to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes. It turns out that the larger the water subcooling is, the more initiation of injection is delayed, and the sparger and the natural circulation of the hot water from the steam generator accelerate the gravity-driven injection. The condensation modes are divided into three modes: sonic jet, subsonic jet, and steam cavity. RELAP5/MOD3.1 is chosen to evaluate the cod predictability on the direct contact condensation in the CMT. It is found that the predictions of MOD3.1 are in better agreement with the experimental data than those of MOD3.0. From the nodalization study of the test section, the 1-node model shows better agreement with the experimental data than the multi-node models. RELAP5/MOD3.1 identifies the flow regime of the test section as vertical stratification. However, the flow regime observed in the experiment is the subsonic jet with the bubble having the vertical cone shape. To accurately predict the direct contact condensation in the CMT with RELAP5/MOD3.1, it is essential that a new set of the interfacial heat transfer coefficients and a new flow regime map for direct contact condensation in the CMT be developed.

  6. RELAP5 based engineering simulator

    SciTech Connect

    Charlton, T.R.; Laats, E.T.; Burtt, J.D.

    1990-01-01

    The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor II (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nuclear Plant Analyzer (NPA), have recently been improved significantly. This paper will present an overview of the INEL Engineering Simulation Center, and discuss the RELAP5/MOD3 and NPA/MOD1 codes, specifically how they are being used at the INEL Engineering Simulation Center. It will provide an update on the modifications to these two codes and their application to the ATR engineering simulator project, as well as, a discussion on the reactor system representation, control system modeling, two phase flow and heat transfer modeling. It will also discuss how these two codes are providing desktop, stand-alone reactor simulation. 12 refs., 2 figs.

  7. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  8. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    DOE PAGESBeta

    Hu, Po; Wilson, Paul

    2014-01-01

    The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less

  9. Workstation experience with RELAP5

    SciTech Connect

    Miller, C.S.; Wagner, R.J.

    1993-08-01

    One of the major improvements in the progress from RELAP5/MOD2 to RELAP5/MOD3 was the modification to improve portability. The use of the code was thus extended from the 60 and 64 bit work mainframe computers to 32 bit workstations and PC`s. This has taken RELAP5 from a few types of mainframes` environmentally controlled computer rooms to many smaller machines running under several different operating systems, with different compilers, and running on, beside, and under engineers desks.

  10. Assessment of RELAP5-3D for Analysis of Very High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Larry Siefken; Cliff Davis

    2005-10-01

    The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code’s capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values.

  11. SCDAP/RELAP5 independent peer review

    SciTech Connect

    Corradini, M.L.; Dhir, V.K.; Haste, T.J.; Heames, T.J.; Jenks, R.P.; Kelly, J.E.; Khatib-Rahbar, M.; Viskanta, R.

    1993-01-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light-water-reactor coolant systems during severe accidents. The newest version of the code is SCDAP/RELAP5/MOD3. The US Nuclear Regulatory Commission (NRC) decided that there was a need for a broad technical review of the code by recognized experts to determine overall technical adequacy, even though the code is still under development. For this purpose, an eight-member SCDAP/RELAP5 Peer Review Committee was organized, and the outcome of the review should help the NRC prioritize future code-development activity. Because the code is designed to be mechanistic, the Committee used a higher standard for technical adequacy than was employed in the peer review of the parametric MELCOR code. The Committee completed its review of the SCDAP/RELAP5 code, and the findings are documented in this report. Based on these findings, recommendations in five areas are provided: (1) phenomenological models, (2) code-design objectives, (3) code-targeted applications, (4) other findings, and (5) additional recommendations.

  12. COBRA/RELAP5; A merged version of the COBRA-TF and RELAP5/MOD3 codes

    SciTech Connect

    Lee, S.Y.; Jeong, J.J.; Kim, S.H. ); Chang, S.H.

    1992-08-01

    This paper reports that the best-estimate thermal-hydraulic codes RELAP5/MOD3 and COBRA-TF were adopted to the Apollo DN 10000 workstation and subsequently merged. This was done to combine the excellent features of the two codes and thus product a code with much enhanced capability. The resulting code was named COBRA/RELAP5. This code has features in common with COBRA/TRAC or TRAC-PF1; three-dimensional reactor vessel and one-dimensional loop modeling capability. The merging of the two codes is focused on the hydrodynamic model and numerical solution schemes. In COBRA/RELAP5, the system pressure matrices of the two codes are merged and solved simultaneously. The merged COBRA/RELAP5 calculations are done in process-level parallel mode on the Apollo DN10000 computer with two central processing units.

  13. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    SciTech Connect

    McCann, Larry D.

    2007-01-30

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  14. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    McCann, Larry D.

    2007-01-01

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  15. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    SciTech Connect

    C. B. Davis; C. H. Oh

    2003-08-01

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide.

  16. PHISICS multi-group transport neutronic capabilities for RELAP5

    SciTech Connect

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G.

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  17. SCDAP/RELAP5 code development and assessment

    SciTech Connect

    Allison, C.M.; Hohorst, J.K.

    1996-03-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code`s calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities.

  18. RELAP5/MOD2 Overview and Developmental. Assessment Results from TMl-1 Plant Transient Analysis

    SciTech Connect

    Lin, J. C.; Tsai, C. C.; Ransom, V. H.; Johnsen, G. W.

    2013-02-01

    RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. The objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly sub-cooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2. (author)

  19. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    NASA Astrophysics Data System (ADS)

    Giardina, M.; Castiglia, F.; Buffa, P.; Palermo, G.; Prete, G.

    2014-11-01

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions.

  20. Molten Salt Mixture Properties (KF-ZrF4 and KCl-MgCl2) for Use in RELAP5-3D for High Temperature Reactor Application

    SciTech Connect

    N. A. Anderson; P. Sabharwall

    2012-06-01

    Molten salt coolants are being investigated as primary coolants for a fluoride high-temperature reactor and as secondary coolants for high temperature reactors such as the next generation nuclear plant. This work provides a review of the thermophysical properties of candidate molten salt coolants for use as a secondary heat transfer medium from a high temperature reactor to a processing plant. The molten salts LiF-NaF-KF, KF-ZrF4 and KCl-MgCl2 were considered for use in the secondary coolant loop. The thermophysical properties necessary to add the molten salts KF-ZrF4 and KCl-MgCl2 to RELAP5-3D were gathered for potential modeling purposes. The properties of the molten salt LiF-NaF-KF were already available in RELAP5-3D. The effect that the uncertainty in individual properties had on the Nusselt number was evaluated. This uncertainty in the Nusselt number was shown to be nearly independent of the molten salt temperature.

  1. RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

    SciTech Connect

    Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

    1990-05-01

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

  2. Modeling the GFR with RELAP5-3D

    SciTech Connect

    Cliff B. Davis; Theron D. Marshall; K. D. Weaver

    2005-09-01

    Significant improvements have been made to the RELAP5-3D computer code for analysis of the Gas Fast Reactor (GFR). These improvements consisted of adding carbon dioxide as a working fluid, improving the turbine component, developing a compressor model, and adding the Gnielinski heat transfer correlation. The code improvements were validated, generally through comparisons with independent design calculations. A model of the power conversion unit of the GFR was developed. The model of the power conversion unit was coupled to a reactor model to develop a complete model of the GFR system. The RELAP5 model of the GFR was used to simulate two transients, one initiated by a reactor trip and the other initiated by a loss of load.

  3. SCDAP/RELAP5/MOD3 code development

    SciTech Connect

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding.

  4. SCDAP/RELAP5/MOD3 code development

    SciTech Connect

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-12-31

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding.

  5. RELAP5YA simulation of large-break tests in the two-loop test apparatus

    SciTech Connect

    da Silva, H.C. Jr.; Sundaram, R.K.; Fernandez, R.T.

    1987-01-01

    The RELAP5YA computer code has been used to simulate two tests on the two-loop test apparatus (TLTA): 6426/1 and 6425/2. The TLTA is a scaled version of a boiling water reactor (BWR). TLTA 6426/1 simulates a double-ended guillotine loss-of-coolant accident (LOCA) in the recirculation line without emergency core cooling (ECC). TLTA 6425/2 is similar except that all ECC systems are operational. RELAP5YA was developed at Yankee Atomic Electric Co. from RELAP5/MOD1 to provide a conservative method for licensing analyses that would also simulate LOCA phenomena in a qualitatively correct manner. Major features distinguishing RELAP5YA from RELAP5/MOD1 include a jet pump component model and constitutive models for interphase drag, critical heat flux, rewet and quench phenomena, and multiple-surface radiation heat transfer. In addition, RELAP5YA contains evaluation model features suitable for performing licensing analyses of light water reactors that conform to regulatory requirements. These evaluation model features have not been used in the present calculations.

  6. RELAP5/MOD3 code manual. Volume 4, Models and correlations

    SciTech Connect

    1995-08-01

    The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations.

  7. Uncertainty Analysis of RELAP5-3D

    SciTech Connect

    Alexandra E Gertman; Dr. George L Mesina

    2012-07-01

    As world-wide energy consumption continues to increase, so does the demand for the use of alternative energy sources, such as Nuclear Energy. Nuclear Power Plants currently supply over 370 gigawatts of electricity, and more than 60 new nuclear reactors have been commissioned by 15 different countries. The primary concern for Nuclear Power Plant operation and lisencing has been safety. The safety of the operation of Nuclear Power Plants is no simple matter- it involves the training of operators, design of the reactor, as well as equipment and design upgrades throughout the lifetime of the reactor, etc. To safely design, operate, and understand nuclear power plants, industry and government alike have relied upon the use of best-estimate simulation codes, which allow for an accurate model of any given plant to be created with well-defined margins of safety. The most widely used of these best-estimate simulation codes in the Nuclear Power industry is RELAP5-3D. Our project focused on improving the modeling capabilities of RELAP5-3D by developing uncertainty estimates for its calculations. This work involved analyzing high, medium, and low ranked phenomena from an INL PIRT on a small break Loss-Of-Coolant Accident as wall as an analysis of a large break Loss-Of- Coolant Accident. Statistical analyses were performed using correlation coefficients. To perform the studies, computer programs were written that modify a template RELAP5 input deck to produce one deck for each combination of key input parameters. Python scripting enabled the running of the generated input files with RELAP5-3D on INL’s massively parallel cluster system. Data from the studies was collected and analyzed with SAS. A summary of the results of our studies are presented.

  8. Preliminary design report for SCDAP/RELAP5 lower core plate model

    SciTech Connect

    Coryell, E.W.; Griffin, F.P.

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a

  9. RBMK thermohydraulic safety assessments using RELAP5/MOD3 codes

    SciTech Connect

    Tsiklauri, G.V.; Schmitt, B.E.

    1995-06-01

    The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to in core pressure tube ruptures. These assessments show that the RELAP5/MOD3 code can predict major phenomena during postulated accidents in the RBMK reactors.

  10. Design report on SCDAP/RELAP5 model improvements - debris bed and molten pool behavior

    SciTech Connect

    Allison, C.M.; Rempe, J.L.; Chavez, S.A.

    1994-11-01

    the SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and in combination with VICTORIA, fission product release and transport during severe accidents. Improvements for existing debris bed and molten pool models in the SCDAP/RELAP5/MOD3.1 code are described in this report. Model improvements to address (a) debris bed formation, heating, and melting; (b) molten pool formation and growth; and (c) molten pool crust failure are discussed. Relevant data, existing models, proposed modeling changes, and the anticipated impact of the changes are discussed. Recommendations for the assessment of improved models are provided.

  11. Uncertainty Analysis for RELAP5-3D

    SciTech Connect

    Aaron J. Pawel; Dr. George L. Mesina

    2011-08-01

    In its current state, RELAP5-3D is a 'best-estimate' code; it is one of our most reliable programs for modeling what occurs within reactor systems in transients from given initial conditions. This code, however, remains an estimator. A statistical analysis has been performed that begins to lay the foundation for a full uncertainty analysis. By varying the inputs over assumed probability density functions, the output parameters were shown to vary. Using such statistical tools as means, variances, and tolerance intervals, a picture of how uncertain the results are based on the uncertainty of the inputs has been obtained.

  12. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    SciTech Connect

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  13. SCDAP/RELAP5/MOD2 code manual

    SciTech Connect

    Allison, C.M.; Johnson, E.C.; Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. )

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs.

  14. RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

    SciTech Connect

    Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

    2005-09-15

    Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test.

  15. SCDAP/RELAP5/MOD2 code manual

    SciTech Connect

    Allison, C.M.; Johnson, E.C.; Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. )

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code.

  16. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    SciTech Connect

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C.

    1995-09-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {plus_minus}5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

  17. Streamlining of the RELAP5-3D Code

    SciTech Connect

    Mesina, George L; Hykes, Joshua; Guillen, Donna Post

    2007-11-01

    RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements have been made decrease code run time and increase solution accuracy. These include vectorization, parallelization, use of improved equation solvers for thermal hydraulics and neutron kinetics, and incorporation of improved library utilities. In the area of applied nuclear engineering, expanded capabilities include boron and level tracking models, radiation/conduction enclosure model, feedwater heater and compressor components, fluids and corresponding correlations for modeling Generation IV reactor designs, and coupling to computational fluid dynamics solvers. Ongoing and proposed future developments include improvements to the two-phase pump model, conversion to FORTRAN 90, and coupling to more computer programs. This paper summarizes the general improvements made to RELAP5-3D, with an emphasis on streamlining the code infrastructure for improved maintenance and development. With all these past, present and planned developments, it is necessary to modify the code infrastructure to incorporate modifications in a consistent and maintainable manner. Modifying a complex code such as RELAP5-3D to incorporate new models, upgrade numerics, and optimize existing code becomes more difficult as the code grows larger. The difficulty of this as well as the chance of introducing errors is significantly reduced when the code is structured. To streamline the code into a structured program, a commercial restructuring tool, FOR_STRUCT, was applied to the RELAP5-3D source files. The

  18. Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results

    SciTech Connect

    Ruggles, A.E.; Morris, D.G.

    1989-01-01

    The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.

  19. AUTOMATED, HIGHLY ACCURATE VERIFICATION OF RELAP5-3D

    SciTech Connect

    George L Mesina; David Aumiller; Francis Buschman

    2014-07-01

    Computer programs that analyze light water reactor safety solve complex systems of governing, closure and special process equations to model the underlying physics. In addition, these programs incorporate many other features and are quite large. RELAP5-3D[1] has over 300,000 lines of coding for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. Verification ensures that a program is built right by checking that it meets its design specifications. Recently, there has been an increased importance on the development of automated verification processes that compare coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions[2]. For the first time, the ability exists to ensure that the data transfer operations associated with timestep advancement/repeating and writing/reading a solution to a file have no unintended consequences. To ensure that the code performs as intended over its extensive list of applications, an automated and highly accurate verification method has been modified and applied to RELAP5-3D. Furthermore, mathematical analysis of the adequacy of the checks used in the comparisons is provided.

  20. Comparison of the PARET/ANL and RELAP5/MOD3 codes for the analysis of IAEA benchmark transients and the SPERT experiments

    SciTech Connect

    Woodruff, W.L.; Hanan, N.A.; Smith, R.S.; Matos, J.E.

    1997-12-01

    The RELAP5/MOD3 code is a coupled kinetics-hydrodynamics code for modelling all components of pressurized water reactor systems. To our knowledge, RELAP5 has not been tested against the SPERT reactivity insertion experiments or more conventional research reactor models such as the 10-MW low-enriched uranium (LEU) benchmark reactor in the International Atomic Energy Agency (IAEA) Guidebook, where loss-of-flow (LOF) and reactivity insertion transients were computed by laboratories in four countries, including Argonne National Laboratory (ANL). The ANL computations used the PARET/ANL code, which has been used extensively for research reactor analysis and compared with the SPERT-I and SPERT-II experiments. RELAP5/MOD3 and PARET/ANL results are compared in this paper. Attempts to compare RELAP/MOD3 with the SPERT experiments are included.

  1. RELAP5-3D code validation for RBMK phenomena

    SciTech Connect

    Fisher, J.E.

    1999-09-01

    The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena.

  2. RELAP5-3D Code Validation for RBMK Phenomena

    SciTech Connect

    Fisher, James Ebberly

    1999-09-01

    The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena.

  3. SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

    SciTech Connect

    Coryell, E.W.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models which are resident in the RELAP5 portion of the code. A description of the organization and structure of SCDAP/RELAP5 is presented. Additional information is provided regarding the manner in which models in one portion of the code impact other parts of the code, and models which are dependent on and derive information from other subcodes.

  4. RELAP5/MOD3 code manual: User`s guide and input requirements. Volume 2

    SciTech Connect

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation.

  5. RELAP5/MOD2 models and correlations

    SciTech Connect

    Dimenna, R.A.; Larson, J.R.; Johnson, R.W.; Larson, T.K.; Miller, C.S.; Streit, J.E.; Hanson, R.G.; Kiser, D.M.

    1988-08-01

    A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included.

  6. A generic semi-implicit coupling methodology for use in RELAP5-3D{copyright}

    SciTech Connect

    Aumiller, D.L.; Tomlinson, E.T.; Weaver, W.L.

    2000-09-01

    A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3D{copyright} computer program. This methodology allows RELAP5-3D{copyright} to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered. The methodology was demonstrated using a test case in which the test geometry was divided into two parts each of which was solved as a RELAP5-3D{copyright} simulation. This test problem exercised all of the semi-implicit coupling features which were installed in RELAP5-3D0. The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process.

  7. RESTRUCTURING RELAP5-3D FOR NEXT GENERATION NUCLEAR PLANT ANALYSIS

    SciTech Connect

    Donna Post Guillen; George L. Mesina; Joshua M. Hykes

    2006-06-01

    RELAP5-3D is used worldwide for analyzing nuclear reactors under both operational transients and postulated accident conditions. Development of the RELAP code series began in 1975 and since that time the code has been continuously improved, enhanced, verified and validated [1]. Since RELAP5-3D will continue to be the premier thermal hydraulics tool well into the future, it is necessary to modernize the code to accommodate the incorporation of additional capabilities to support the development of the next generation of nuclear reactors [2]. This paper discusses the reengineering of RELAP5-3D into structured code.

  8. Modifications to the VV PHTS RELAP5 Model

    SciTech Connect

    Carbajo, Juan J

    2011-02-01

    Modifications and improvements to a previous RELAP5 model of the Vacuum Vessel Primary Heat Transfer System are described in this report. The modifications were new pump models, a new steam pressurizer, new coolant water control systems, additional pipe structures, and reduction of the pulse power to 6 MW.

  9. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    NASA Astrophysics Data System (ADS)

    Sabundjian, Gaianê; Andrade, Delvonei A.; Belchior, Antonio, Jr.; da Silva Rocha, Marcelo; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; de Souza Lima, Ana Cecília

    2013-05-01

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  10. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    SciTech Connect

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C.; Rohatgi, U.S.

    1995-09-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations al these conditions were compared with the GIRAFFE data. The effects of PCCS cell nodings on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {+-}5% of the data with a three-node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer in the presence of noncondensable gases with only a coarse mesh. The cell length term in the condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

  11. RELAP5-3D Developer Guidelines and Programming Practices

    SciTech Connect

    Dr. George L Mesina

    2014-03-01

    Our ultimate goal is to create and maintain RELAP5-3D as the best software tool available to analyze nuclear power plants. This begins with writing excellent programming and requires thorough testing. This document covers development of RELAP5-3D software, the behavior of the RELAP5-3D program that must be maintained, and code testing. RELAP5-3D must perform in a manner consistent with previous code versions with backward compatibility for the sake of the users. Thus file operations, code termination, input and output must remain consistent in form and content while adding appropriate new files, input and output as new features are developed. As computer hardware, operating systems, and other software change, RELAP5-3D must adapt and maintain performance. The code must be thoroughly tested to ensure that it continues to perform robustly on the supported platforms. The coding must be written in a consistent manner that makes the program easy to read to reduce the time and cost of development, maintenance and error resolution. The programming guidelines presented her are intended to institutionalize a consistent way of writing FORTRAN code for the RELAP5-3D computer program that will minimize errors and rework. A common format and organization of program units creates a unifying look and feel to the code. This in turn increases readability and reduces time required for maintenance, development and debugging. It also aids new programmers in reading and understanding the program. Therefore, when undertaking development of the RELAP5-3D computer program, the programmer must write computer code that follows these guidelines. This set of programming guidelines creates a framework of good programming practices, such as initialization, structured programming, and vector-friendly coding. It sets out formatting rules for lines of code, such as indentation, capitalization, spacing, etc. It creates limits on program units, such as subprograms, functions, and modules. It

  12. Conservation of Fluid Mass and Energy by RELAP5-3D during a SBLOCA

    SciTech Connect

    Cliff B. Davis

    2009-08-01

    Mass and energy balances were performed to check the accuracy of RELAP5-3D’s solution during a loss-of-coolant accident initiated by a small break in a typical pressurized water reactor. Mass and energy balances were performed for the combined liquid and gas phases and the gas phase by itself. The analysis showed that RELAP5-3D adequately conserved mass and energy for the combined fluid and the gas phase.

  13. Interpretation of TRIGA reactivity transients with RELAP5/PARCS coupled-code

    SciTech Connect

    Bandini, G.; Meloni, P.; Polidori, M.

    2006-07-01

    In the frame of future experiments to carried out upon TRIGA reactors, which aim to verify the real feasibility of the ADS (Accelerator Driven System) concept, it is essential to build a numerical tool able to simulate the dynamic behaviour of the reactor in subcritical configuration. This model developed to support the design of subcritical experiments and the safety analysis of the reactor, as a first step has to be assessed against the experimental data available for the critical reactor. To this purpose the thermal-hydraulic/ neutronic numerical model based on the RELAP5/PARCS coupled-code is been tested against the experimental reactivity transients conducted on the RC1-TRIGA reactor at the ENEA Casaccia Research Center in forecast of the TRADE (TRIGA Accelerator Driven Experiment) subcritical experience. The results of the calculations already performed show a qualitative good agreement with the experimental data and allow to address the future developments and improvements of the numerical model. (authors)

  14. SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

    SciTech Connect

    Hohorst, J.K.; Johnsen, E.C.; Allison, C.M.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported.

  15. Application of RELAP5/MOD3.1 code to the LOFT test L3-6

    SciTech Connect

    Pylev, S.S.; Roginskaja, V.L.

    1998-02-01

    A calculation of LOFT Experiment L3-6, a small break equivalent to a 4-in diameter rupture in the cold leg of a four-loop commercial pressurized water reactor, has been performed to help validate RELAP5/MOD3.1 for this application. The version of the code to be used is SCDAP/RELAP5/MOD3.1.8d0. Three calculations were carried out in order to study the sensitivity to change break nozzle superheated discharge coefficient. Conducted comparative analysis of the LOFT L3-6 experiment shows on the whole a reasonable agreement between calculated data. Some discrepancies in the system pressure do not distort a picture of the transient. 6 refs.

  16. RELAP5-3D Resolution of Known Restart/Backup Issues

    SciTech Connect

    Mesina, George L.; Anderson, Nolan A.

    2014-12-01

    The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

  17. Analysis of panthers full-scale heat transfer tests with RELAP5

    SciTech Connect

    Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

    1996-01-01

    The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric`s (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit.

  18. Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor using RELAP5 and TEMPEST: Part 1, Models and simulation results

    SciTech Connect

    Morris, D.G.; Wendel, M.W.; Chen, N.C.J.; Ruggles, A.E.; Cook, D.H.

    1989-01-01

    A study was conducted to examine decay heat removal requirements in the High Flux Isotope Reactor (HFIR) following shutdown from 85 MW. The objective of the study was to determine when forced flow through the core could be terminated without causing the fuel to melt. This question is particularly relevant when a station blackout caused by an external event is considered. Analysis of natural circulation in the core, vessel upper plenum, and reactor pool indicates that 12 h of forced flow will permit a safe shutdown with some margin. However, uncertainties in the analysis preclude conclusive proof that 12 h is sufficient. As a result of the study, two seismically qualified diesel generators were installed in HFIR. 9 refs., 4 figs.

  19. Extremely accurate sequential verification of RELAP5-3D

    DOE PAGESBeta

    Mesina, George L.; Aumiller, David L.; Buschman, Francis X.

    2015-11-19

    Large computer programs like RELAP5-3D solve complex systems of governing, closure and special process equations to model the underlying physics of nuclear power plants. Further, these programs incorporate many other features for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. For RELAP5-3D, verification and validation are restricted to nuclear power plant applications. Verification means ensuring that the program is built right by checking that it meets its design specifications, comparing coding to algorithms and equations and comparing calculations against analytical solutions and method ofmore » manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. Recently, an automated, highly accurate sequential verification method has been developed for RELAP5-3D. The method also provides to test that no unintended consequences result from code development in the following code capabilities: repeating a timestep advancement, continuing a run from a restart file, multiple cases in a single code execution, and modes of coupled/uncoupled operation. In conclusion, mathematical analyses of the adequacy of the checks used in the comparisons are provided.« less

  20. Extremely accurate sequential verification of RELAP5-3D

    SciTech Connect

    Mesina, George L.; Aumiller, David L.; Buschman, Francis X.

    2015-11-19

    Large computer programs like RELAP5-3D solve complex systems of governing, closure and special process equations to model the underlying physics of nuclear power plants. Further, these programs incorporate many other features for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. For RELAP5-3D, verification and validation are restricted to nuclear power plant applications. Verification means ensuring that the program is built right by checking that it meets its design specifications, comparing coding to algorithms and equations and comparing calculations against analytical solutions and method of manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. Recently, an automated, highly accurate sequential verification method has been developed for RELAP5-3D. The method also provides to test that no unintended consequences result from code development in the following code capabilities: repeating a timestep advancement, continuing a run from a restart file, multiple cases in a single code execution, and modes of coupled/uncoupled operation. In conclusion, mathematical analyses of the adequacy of the checks used in the comparisons are provided.

  1. RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

    SciTech Connect

    Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S.

    1995-09-01

    Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

  2. Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

    SciTech Connect

    Park, R.J.; Kim, S.B.; Kim, H.D.

    2004-07-01

    High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

  3. RELAP5 assessment: LOFT intermediate breaks L5-1 and L8-2. [PWR

    SciTech Connect

    Orman, J.L.; Kmetyk, L.N.

    1983-08-01

    The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The RELAP5 code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, an intermediate break transient (L5-1) and a core uncovery transient (L8-2) performed at the LOFT facility have been analyzed. Transient calculations were done with both cycle 14 and cycle 18 of RELAP5/MOD1 and a comparison of the predictions was made. The results show that RELAP5/MOD1 did very well calculating the overall behavior for these intermediate break experiments, although there were a few quantitative disagreements.

  4. User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

    SciTech Connect

    Forsmann, J. Hope; Weaver, Walter L.

    2015-04-01

    This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

  5. Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

    SciTech Connect

    Aragones, Jose M.; Ahnert, Carol; Cabellos, Oscar; Garcia-Herranz, Nuria; Aragones-Ahnert, Vanessa

    2004-04-15

    The purpose of this paper is first to discuss the methods developed in our three-dimensional pressurized water reactor core dynamics code SIMTRAN and its coupling to the system code RELAP-5 for general transient and safety analysis. Then, we summarize its demonstration application to the Nuclear Energy Agency (NEA)/Organization for Economic Cooperation and Development (OECD) Benchmark on Main Steam Line Break (MSLB), co-sponsored by the U.S. Nuclear Regulatory Commission (NRC) and other regulatory institutions. In particular, our work has been supported by the Spanish 'Consejo de Seguridad Nuclear' (CSN) under a CSN research project.Our results for the steady states and the guided-core transients, proposed as exercise 2 of the MSLB benchmark, show small deviations from the mean results of all participants, especially in core average parameters. For the full-coupled core-plant transients, exercise 3, a detailed comparison with the University of Purdue-NRC results using PARCS/RELAP-5, shows quite good agreement in both integral and local parameters, especially for the more extreme return-to-power scenario.

  6. Restructuring of RELAP5-3D

    SciTech Connect

    George Mesina; Joshua Hykes

    2005-09-01

    The RELAP5-3D source code is unstructured with many interwoven logic flow paths. By restructuring the code, it becomes easier to read and understand, which reduces the time and money required for code development, debugging, and maintenance. A structured program is comprised of blocks of code with one entry and exit point and downward logic flow. IF tests and DO loops inherently create structured code, while GOTO statements are the main cause of unstructured code. FOR_STRUCT is a commercial software package that converts unstructured FORTRAN into structured programming; it was used to restructure individual subroutines. Primarily it transforms GOTO statements, ARITHMETIC IF statements, and COMPUTED GOTO statements into IF-ELSEIF-ELSE tests and DO loops. The complexity of RELAP5-3D complicated the task. First, FOR_STRUCT cannot completely restructure all the complex coding contained in RELAP5-3D. An iterative approach of multiple FOR_STRUCT applications gave some additional improvements. Second, FOR_STRUCT cannot restructure FORTRAN 90 coding, and RELAP5-3D is partially written in FORTRAN 90. Unix scripts for pre-processing subroutines into coding that FOR_STRUCT could handle and post-processing it back into FORTRAN 90 were written. Finally, FOR_STRUCT does not have the ability to restructure the RELAP5-3D code which contains pre-compiler directives. Variations of a file were processed with different pre-compiler options switched on or off, ensuring that every block of code was restructured. Then the variations were recombined to create a completely restructured source file. Unix scripts were written to perform these tasks, as well as to make some minor formatting improvements. In total, 447 files comprising some 180,000 lines of FORTRAN code were restructured. These showed significant reduction in the number of logic jumps contained as measured by reduction in the number of GOTO statements and line labels. The average number of GOTO statements per subroutine

  7. FPTRAN: A Volatile Fission Products and Structural Materials Transport Code for SCDAP/RELAP5

    SciTech Connect

    Honaiser, Eduardo; Anghaie, Samim

    2004-07-01

    The fission products behavior in reactor coolant systems (RCS) is divided in the fission products release from the fuel, transport through the piping system, and the chemistry of the several materials present in a LWR. The transport poses significant difficulty for the implementation, due to the complexity in the treatment of the system of equations generated for the solution, as well as the difficulties in the modeling of certain phenomena. This paper presents the FPTRAN code, which was incorporated to SCDAP/RELAP5, and initially tested satisfactorily. FPTRAN does the calculation of the transport of fission products in RCS, estimating the amount of material being deposited over the pipes, and the amount released to the containment, once a source of released material (fission products and structural materials) to the piping system is provided. (authors)

  8. RELAP5 assessment: LOFT large break L2-5

    SciTech Connect

    Thompson, S L; Kmetyk, L N

    1984-02-01

    RELAP5 is part of an effort to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The RELAP5 code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, a large break transient performed at the LOFT facility has been analyzed. The results show that RELAP5/MOD1 correctly calculates many of the major system variables (i.e., pressure, break flows, peak clad temperature) early in a large break LOCA. The major problems encountered in the analyses were incorrect pump coastdown and loop seal clearing early in the calculation, excessive pump speedup later in the transient (probably due to too much condensation-induced pressure drop at the ECC injection point), and excess ECC bypass calculated throughout the later portions of the test; only the latter problem significantly affected the overall results. This excess ECC bypass through the downcomer and vessel-side break resulted in too-large late-time break flows and high system pressure due to prolonged choked flow conditions. It also resulted in a second core heatup being calculated after the accumulator emptied, since water was not being retained in the vessel. Analogous calculations with a split-downcomer nodalization delivered some ECC water to the lower plenum, which was then swept up the core and upper plenum and out the other (pump-side) break; thus no significant differences in long-term overall behavior were evident between the calculations.

  9. RHF RELAP5 Model and Preliminary Loss-Of-Offsite-Power Simulation Results for LEU Conversion

    SciTech Connect

    Licht, J. R.; Bergeron, A.; Dionne, B.; Thomas, F.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  10. A comparison of the PARET/ANL and RELAP5/MOD3 codes for the analysis of IAEA benchmark transients

    SciTech Connect

    Woodruff, W.L.; Hanan, N.A.; Smith, R.S.; Matos, J.E.

    1996-12-31

    The PARET/ANL and RELAP5/MOD3 codes are used to analyze the series of benchmark transients specified for the IAEA Research Reactor Core Conversion Guidebook (IAEA-TECDOC-643, Vol. 3). The computed results for these loss-of-flow and reactivity insertion transients with scram are in excellent agreement and agree well with the earlier results reported in the guidebook. Attempts to also compare RELAP5/MOD3 with the SPERT series of experiments are in progress.

  11. The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

    SciTech Connect

    Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior, Antonio Jr.; Silva Rocha, Marcelo da; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; Souza Lima, Ana Cecilia de

    2013-05-06

    This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm{sup 2}, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

  12. RELAP5-3D Restart and Backup Verification Testing

    SciTech Connect

    Dr. George L Mesina

    2013-09-01

    Existing testing methodology for RELAP5-3D employs a set of test cases collected over two decades to test a variety of code features and run on a Linux or Windows platform. However, this set has numerous deficiencies in terms of code coverage, detail of comparison, running time, and testing fidelity of RELAP5-3D restart and backup capabilities. The test suite covers less than three quarters of the lines of code in the relap directory and just over half those in the environmental library. Even in terms of code features, many are not covered. Moreover, the test set runs many problems long past the point necessary to test the relevant features. It requires standard problems to run to completion. This is unnecessary for features can be tested in a short-running problem. For example, many trips and controls can be tested in the first few time steps, as can a number of fluid flow options. The testing system is also inaccurate. For the past decade, the diffem script has been the primary tool for checking that printouts from two different RELAP5-3D executables agree. This tool compares two output files to verify that all characters are the same except for those relating to date, time and a few other excluded items. The variable values printed on the output file are accurate to no more than eight decimal places. Therefore, calculations with errors in decimal places beyond those printed remain undetected. Finally, fidelity of restart is not tested except in the PVM sub-suite and backup is not specifically tested at all. When a restart is made from any midway point of the base-case transient, the restart must produce the same values. When a backup condition occurs, the code repeats advancements with the same time step. A perfect backup can be tested by forcing RELAP5 to perform a backup by falsely setting a backup condition flag at a user-specified-time. Comparison of the calculations of that run and those produced by the same input w/o the spurious condition should be

  13. Extensions to SCDAP/RELAP5/ATHENA for Analysis of HTGRs and SCWRs

    SciTech Connect

    E. A. Harvego; L. J. Siefken

    2004-04-01

    The SCDAP/RELAP5/ATHENA code was extended to enable the code to perform transient analyses of High Temperature Gas Reactors (HTGRs). Preliminary results indicate that post-shutdown decay heat can be adequately removed from HTGRs by natural circulation of atmospheric air.

  14. Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

    SciTech Connect

    Aragones, Jose M.; Ahnert, Carol; Cabellos, Oscar; Garcia-Herranz, Nuria

    2001-06-17

    This work discusses the methods developed in a three-dimensional (3-D) pressurized water reactor (PWR) SIMTRAN Core Dynamics code and its coupling to the RELAP-5 system code for general transient and safety analysis, as well as its demonstration application to the Nuclear Energy Agency/Organization for Economic Cooperation and Development (NEA/OECD) Benchmark on Main Steam Line Break (MSLB). The results for the steady states and the transients proposed for exercise 2 of the MSLB Benchmark, the guided core transient analysis, show small deviations from the mean results of other participants, especially for core average parameters. For exercise 3, the full system transient, the agreement is quite good for both integral and local parameters.

  15. Evaluation of RELAP5 MOD 3.1.1 code with GIRAFFE Test Facility: Phase 1, Step 2 nitrogen venting tests

    SciTech Connect

    Boyer, B.D.; Slovik, G.C.; Rohatgl, U.S.

    1995-11-01

    The Simplified Boiling Water Reactor (SBWR) proposed by General Electric (GE) is an advanced light water reactor (ALWR) design that utilizes passive safety systems. The PCCS is a series of heat exchangers submerged in water and open to the containment. Since the containment is inerted with nitrogen during normal operation, the PCCS must condense the steam in the presence of noncondensable gases during an accident. To model the transient behavior of the SBWR with a system code, the code should properly simulate the expected phenomena. To validate the applicability of RELAP5 MOD 3.1.1, the data from three Phase 1, Step 2 nitrogen venting tests at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal facility and RELAP5 calculations of these tests were compared. The comparison of the GIRAFFE data against the results from the RELAP5 calculations showed that it can predict condensation and gas purging phenomena occurring in the long-term decay heat rejection phase. In this phase of the transient, condensation in the PCCS is the only means to reject heat from the SBWR containment. In the two tests where the nitrogen purge vent line was at its deepest submergence in the Suppression Pool (SIP), the RELAP5 results mirrored the behavior of the containment pressures and of the water levels in the Horizontal Vent (HV) and the nitrogen purge line tube of the GIRAFFE data. However, in the test with the shallowest purge line submergence, there was appreciable direct contact condensation on the pool surface of the HV despite modeling efforts to deter these phenomena. This surface condensation, unobserved in the GIRAFFE tests, was a major cause of RELAP5 predicting early containment depressurization and the subsequent early rise in HV and nitrogen purge line water levels. The present RELAP5 MOD3.1.1 interfacial heat and mass transfer model does not properly degrade direct contact steam condensation in the presence of noncondensable gases sitting on a pool.

  16. Multi-componenet diffusion analysis and assessment of Gamma code and improved RELAP5 code

    SciTech Connect

    Chang Oh

    2007-05-01

    A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature. New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary). A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel and the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.

  17. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    SciTech Connect

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

  18. RELAP5-3D Code Includes Athena Features and Models

    SciTech Connect

    Richard A. Riemke; Cliff B. Davis; Richard R. Schultz

    2006-07-01

    Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, sf6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5- 3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper.

  19. RELAP5-3D Code Includes ATHENA Features and Models

    SciTech Connect

    Riemke, Richard A.; Davis, Cliff B.; Schultz, Richard R.

    2006-07-01

    Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF{sub 6}, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)

  20. PUMA-PCCS separate effect tests and RELAP5 code evaluation in PUMA

    NASA Astrophysics Data System (ADS)

    Choi, Sung Won

    One of the key areas in the design of advanced nuclear reactors is to develop a reliable Passive Containment Cooling System (PCCS). The purpose of the current work is to better understand the condensation phenomena in PCCS for the downward co-current flow of a steam/air mixture through condenser tube bundles during the three PCCS operational modes, namely the bypass mode, the cyclic venting mode and the long-term cooling mode. A series of unique separate-effect PCCS test data were obtained for condensation heat transfer in the PCCS heat exchangers of the PUMA (Purdue University Multidimensional Integral Test Assembly) facility under a task sponsored by the U.S. Nuclear Regulatory Commission. Test conditions includes bypass mode, cyclic venting mode and long term mode, covering a wide range of Loss of Coolant Accident(LOCA) conditions with a parameters of pressure, mass flow rate, noncondensable(NC) gases, and PCCS pool water level. The parametric effect studies and a further validation of the PUMA-PCCS separate effect test data were performed. The evaluation of a best estimate system code (RELAP5/MOD3.3) was performed by using unique PUMA-PCCS separate effects data and PUMA-Main Steam Line Break (MSLB) integral test (1998). Through a sensitivity studies of nodalization method and physical models on the MSLB test simulations, deficiencies in RELAP5/MOD3.3 code were found as follows: (1) over prediction of heat removal rate by condensation models, (2) overestimation of SP heat transfer through the horizontal venting line and thermal stratification distortion, (3) underestimation of NC gas effects in PCCS by the distortion of cyclic venting phenomena and (4) overestimation of the DW and SP wall condensation. The improvement for the code calculation predictions could be obtained by removing the RELAP5/MOD3.3 code deficient factors in the PUMA MSLB integral test simulation. The unique PCCS NC gas venting visualizations were obtained according to various PCCS inlet NC

  1. Transient validation of RELAP5 model with the DISS facility in once through operation mode

    NASA Astrophysics Data System (ADS)

    Serrano-Aguilera, J. J.; Valenzuela, L.

    2016-05-01

    Thermal-hydraulic code RELAP5 has been used to model a Solar Direct Steam Generation (DSG) system. Experimental data from the DISS facility located at Plataforma Solar de Almería is compared to the numerical results of the RELAP5 model in order to validate it. Both the model and the experimental set-up are in once through operation mode where no injection or active control is regarded. Time dependent boundary conditions are taken into account. This work is a preliminary study of further research that will be carried out in order to achieve a thorough validation of RELAP5 models in the context of DSG in line-focus solar collectors.

  2. RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

    SciTech Connect

    Gerhard Strydom

    2012-06-01

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.

  3. RELAP5-3D results for phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW benchmark

    SciTech Connect

    Strydom, G.; Epiney, A. S.

    2012-07-01

    The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requires participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2. (authors)

  4. Validation and verification of RELAP5 for Advanced Neutron Source accident analysis: Part I, comparisons to ANSDM and PRSDYN codes

    SciTech Connect

    Chen, N.C.J.; Ibn-Khayat, M.; March-Leuba, J.A.; Wendel, M.W.

    1993-12-01

    As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety analysis. Both the ANSDM and PRSDYN models use a simplified single-phase equation set to predict transient thermal-hydraulic performance. Brief descriptions of each of the codes, models, and model limitations were included. Even though comparisons were limited to single-phase conditions, a broad spectrum of accidents was benchmarked: a small loss-of-coolant-accident (LOCA), a large LOCA, a station blackout, and a reactivity insertion accident. The overall conclusion is that the three models yield similar results if the input parameters are the same. However, ANSDM does not capture pressure wave propagation through the coolant system. This difference is significant in very rapid pipe break events. Recommendations are provided for further model improvements.

  5. RELAP5 assessment using semiscale SBLOCA test S-NH-1. International Agreement Report

    SciTech Connect

    Lee, E.J.; Chung, B.D.; Kim, H.J.

    1993-06-01

    2-inch cold leg break test S-NH-1, conducted at the 1/1705 volume scaled facility Semiscale was analyzed using RELAP5/MOD2 Cycle 36.04 and MOD3 Version 5m5. Loss of HPIS was assumed, and reactor trip occurred on a low PZR pressure signal (13.1 MPa), and pumps began an unpowered coastdown on SI signal (12.5 MPa). The system was recovered by opening ADV`s when the PCT became higher than 811 K. Accumulator was finally injected into the system when the primary system pressure was less than 4.0 MPa. The experiment was terminated when the pressure reached the LPIS actuation set point RELAP5/MOD2 analysis demonstrated its capability to predict, with a sufficient accuracy, the main phenomena occurring in the depressurization transient, both from a qualitative and quantitative points of view. Nevertheless, several differences were noted regarding the break flow rate and inventory distribution due to deficiencies in two-phase choked flow model, horizontal stratification interfacial drag, and a CCFL model. The main reason for the core to remain nearly fully covered with the liquid was the under-prediction of the break flow by the code. Several sensitivity calculations were tried using the MOD2 to improve the results by using the different options of break flow modeling (downward, homogeneous, and area increase). The break area compensating concept based on ``the integrated break flow matching`` gave the best results than downward junction and homogeneous options. And the MOD3 showed improvement in predicting a CCFL in SG and a heatup in the core.

  6. SCDAP/RELAP5/MOD 3.1 code manual: User`s guide and input manual. Volume 3

    SciTech Connect

    Coryell, E.W.; Johnsen, E.C.; Allison, C.M.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume provides guidelines to code users based upon lessons learned during the developmental assessment process. A description of problem control and the installation process is included. Appendix a contains the description of the input requirements.

  7. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    SciTech Connect

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  8. Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD MHTGR-350 Benchmark

    SciTech Connect

    Gerhard Strydom

    2014-04-01

    The INL PHISICS code system consists of three modules providing improved core simulation capability: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been finalized, and as part of the code verification and validation program the exercises defined for Phase I of the OECD/NEA MHTGR 350 MW Benchmark were completed. This paper provides an overview of the MHTGR Benchmark, and presents selected results of the three steady state exercises 1-3 defined for Phase I. For Exercise 1, a stand-alone steady-state neutronics solution for an End of Equilibrium Cycle Modular High Temperature Reactor (MHTGR) was calculated with INSTANT, using the provided geometry, material descriptions, and detailed cross-section libraries. Exercise 2 required the modeling of a stand-alone thermal fluids solution. The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 combined the first two exercises in a coupled neutronics and thermal fluids solution, and the coupled code suite PHISICS/RELAP5-3D was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of the traditional RELAP5-3D “ring” model approach vs. a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity of the block model is illustrated with comparison results on the temperature, power density and flux distributions, and the typical under-predictions produced by the ring model approach are highlighted.

  9. RELAP5-3D Compressor Model

    SciTech Connect

    James E. Fisher; Cliff B. Davis; Walter L. Weaver

    2005-06-01

    A compressor model has been implemented in the RELAP5-3D© code. The model is similar to that of the existing pump model, and performs the same function on a gas as the pump performs on a single-phase or two-phase fluid. The compressor component consists of an inlet junction and a control volume, and optionally, an outlet junction. This feature permits cascading compressor components in series. The equations describing the physics of the compressor are derived from first principles. These equations are used to obtain the head, the torque, and the energy dissipation. Compressor performance is specified using a map, specific to the design of the machine, in terms of the ratio of outlet-to-inlet total (or stagnation) pressure and adiabatic efficiency as functions of rotational velocity and flow rate. The input quantities are specified in terms of dimensionless variables, which are corrected to stagnation density and stagnation sound speed. A small correction was formulated for the input of efficiency to account for the error introduced by assumption of constant density when integrating the momentum equation. Comparison of the results of steady-state operation of the compressor model to those of the MIT design calculation showed excellent agreement for both pressure ratio and power.

  10. Plant Transient Analysis for TAPS-3 and 4 using RELAP-5/MOD 3.2

    SciTech Connect

    Sharma, S.L.; Banerjee, S.; Rammohan, H.P.; Malhotra, P.K; Ghadge, S.G.; Bajaj, S.S.

    2006-07-01

    Tarapur Atomic Power Station- 3 and 4 (TAPS-3 and 4) are 540 Mwe Pressurised Heavy Water Reactors (PHWR). This paper presents the results of transient analysis performed to investigate the variations in different parameters of the power plant due to a spurious SETBACK signal which results in power reduction from 100% to 65% FP at a rate of 0.5% FP/sec. The analysis has been done using system thermal hydraulics code RELAP5/MOD3.2. Although the code has been widely used for modeling the Pressurised Water Reactors (PWR); Modeling of Indian PHWRs, which are horizontal pressure tube type reactors, has been done using the code for this analysis. Here the overall thermal hydraulics of the plant as well as various control systems along with trip and actuation logics have been simulated. The variation of different parameters like SG pressure, PHT pressure, SG level, etc. have been studied. An additional study has been done to evaluate the performance of the steam generator at different power levels. (authors)

  11. SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

    SciTech Connect

    Davis, K.L.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made.

  12. An implicit steady-state initialization package for the RELAP5 computer code

    SciTech Connect

    Paulsen, M.P.; Peterson, C.E.; Odar, F.

    1995-08-01

    A direct steady-state initialization (DSSI) method has been developed and implemented in the RELAP5 hydrodynamic analysis program. It provides a means for users to specify a small set of initial conditions which are then propagated through the remainder of the system. The DSSI scheme utilizes the steady-state form of the RELAP5 balance equations for nonequilibrium two-phase flow. It also employs the RELAP5 component models and constitutive model packages for wall-to-phase and interphase momentum and heat exchange. A fully implicit solution of the linearized hydrodynamic equations is implemented. An implicit coupling scheme is used to augment the standard steady-state heat conduction solution for steam generator use. It solves the primary-side tube region energy equations, heat conduction equations, wall heat flux boundary conditions, and overall energy balance equation as a coupled system of equations and improves convergence. The DSSI method for initializing RELAP5 problems to steady-state conditions has been compared with the transient solution scheme using a suite of test problems including; adiabatic single-phase liquid and vapor flow through channels with and without healing and area changes; a heated two-phase test bundle representative of BWR core conditions; and a single-loop PWR model.

  13. Comparison of the PHISICS/RELAP5-3D ring and block model results for phase I of the OECD/NEA MHTGR-350 benchmark

    SciTech Connect

    Strydom, G.; Epiney, A. S.; Alfonsi, Andrea; Rabiti, Cristian

    2015-12-02

    The PHISICS code system has been under development at INL since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and modules performing criticality searches, fuel shuffling and generalized perturbation. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort the first phase of the OECD/NEA MHTGR-350 Benchmark has now been completed. The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detail in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 Benchmark and presents the results of Exercises 2 and 3 defined for Phase I. Exercise 2 required the modelling of a stand-alone thermal fluids solution at End of Equilibrium Cycle for the Modular High Temperature Reactor (MHTGR). The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity that can be obtained by this “block” model is illustrated with comparison results on the temperature, power density and flux distributions. Furthermore, it is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher fidelity block model, and that the additional model development and run-time efforts are worth the gains obtained in the

  14. Comparison of the PHISICS/RELAP5-3D ring and block model results for phase I of the OECD/NEA MHTGR-350 benchmark

    DOE PAGESBeta

    Strydom, G.; Epiney, A. S.; Alfonsi, Andrea; Rabiti, Cristian

    2015-12-02

    The PHISICS code system has been under development at INL since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and modules performing criticality searches, fuel shuffling and generalized perturbation. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort the first phase of the OECD/NEA MHTGR-350 Benchmark has now been completed. The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detailmore » in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 Benchmark and presents the results of Exercises 2 and 3 defined for Phase I. Exercise 2 required the modelling of a stand-alone thermal fluids solution at End of Equilibrium Cycle for the Modular High Temperature Reactor (MHTGR). The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity that can be obtained by this “block” model is illustrated with comparison results on the temperature, power density and flux distributions. Furthermore, it is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher fidelity block model, and that the additional model development and run-time efforts are worth the gains obtained

  15. Blind-blind prediction by RELAP5/MOD1 for a 0. 1% very small cold-leg break experiment at ROSA-IV large-scale test facility

    SciTech Connect

    Koizumi, Y.; Kumamaru, H.; Kukita, Y.; Kawaji, M.; Osakabe, M.; Schultz, R.R.; Tanaka, M.; Tasaka, K.

    1986-06-01

    The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for the analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily.

  16. Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

    SciTech Connect

    Nathan V. Hoffer; Piyush Sabharwall; Nolan A. Anderson

    2011-01-01

    Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.

  17. BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

    SciTech Connect

    Mandelli, D.; Smith, C.; Riley, T.; Nielsen, J.; Alfonsi, A.; Cogliati, J.; Rabiti, C.; Schroeder, J.

    2016-01-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Water Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Lastly, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.

  18. BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

    DOE PAGESBeta

    Mandelli, D.; Smith, C.; Riley, T.; Nielsen, J.; Alfonsi, A.; Cogliati, J.; Rabiti, C.; Schroeder, J.

    2016-01-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Watermore » Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Lastly, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.« less

  19. IMPROVEMENTS TO THE TIME STEPPING ALGORITHM OF RELAP5-3D

    SciTech Connect

    Cumberland, R.; Mesina, G.

    2009-01-01

    The RELAP5-3D time step method is used to perform thermo-hydraulic and neutronic simulations of nuclear reactors and other devices. It discretizes time and space by numerically solving several differential equations. Previously, time step size was controlled by halving or doubling the size of a previous time step. This process caused the code to run slower than it potentially could. In this research project, the RELAP5-3D time step method was modifi ed to allow a new method of changing time steps to improve execution speed and to control error. The new RELAP5-3D time step method being studied involves making the time step proportional to the material courant limit (MCL), while insuring that the time step does not increase by more than a factor of two between advancements. As before, if a step fails or mass error is excessive, the time step is cut in half. To examine performance of the new method, a measure of run time and a measure of error were plotted against a changing MCL proportionality constant (m) in seven test cases. The removal of the upper time step limit produced a small increase in error, but a large decrease in execution time. The best value of m was found to be 0.9. The new algorithm is capable of producing a signifi cant increase in execution speed, with a relatively small increase in mass error. The improvements made are now under consideration for inclusion as a special option in the RELAP5-3D production code.

  20. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    Energy Science and Technology Software Center (ESTSC)

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  1. Architectural Advancements in RELAP5-3D

    SciTech Connect

    Dr. George L. Mesina

    2005-11-01

    As both the computer industry and field of nuclear science and engineering move forward, there is a need to improve the computing tools used in the nuclear industry to keep pace with these changes. By increasing the capability of the codes, the growing modeling needs of nuclear plant analysis will be met and advantage can be taken of more powerful computer languages and architecture. In the past eighteen months, improvements have been made to RELAP5-3D [1] for these reasons. These architectural advances include code restructuring, conversion to Fortran 90, high performance computing upgrades, and rewriting of the RELAP5 Graphical User Interface (RGUI) [2] and XMGR5 [3] in Java. These architectural changes will extend the lifetime of RELAP5-3D, reduce the costs for development and maintenance, and improve it speed and reliability.

  2. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    SciTech Connect

    Banati, J.

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  3. Using an IIST 1% Cold-Leg SBLOCA Experiment with Passive Safety Injection to Assess the RELAP5/MOD3.2 Code

    SciTech Connect

    Lee, C.-H.; Huang, I-M.; Chang, C.-J

    2001-08-15

    The thermal-hydraulic behavior of a postulated 1% cold-leg break loss-of-coolant accident (LOCA) in a pressurized water reactor system was investigated experimentally by the three-loop Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility with the passive core cooling system (PCCS) and numerically by the RELAP5/MOD3.2 computer code. The PCCS of the IIST facility includes three core makeup tanks (CMTs), three accumulators, and a four-stage automatic depressurization system. The aim of this research is to study the performance of the CMTs with the actuation of the ADS during a small-break LOCA. The experimental results show that the IIST PCCS has the capability to maintain long-term cooling under a postulated 1% cold-leg break LOCA. The comparison of the RELAP5/MOD3.2 simulation against the experimental data shows good agreement in major thermal-hydraulic phenomena in the reactor coolant system, but the prediction of the asymmetric behavior for the three CMTs during a gravity drain period is inadequate.

  4. Power loop modeling and simulation using LabVIEW coupled with RELAP5

    NASA Astrophysics Data System (ADS)

    Pack, Joshua C.

    The purpose of this thesis is to provide an additional tool to researchers and system analysts for use in simulation, testing, and development of the secondary loop of a PWR nuclear power plant. This new tool is a coupling of LabVIEW and RELAP5 that has been created by using each code to model half of a PWR. By taking advantage of the strengths of both programs, a more powerful, adaptable, and user friendly system model is developed that links directly to the instrumentation of the system. This work includes the development of the LabVIEW secondary loop model, the coupling methods for linking the two software packages, and a comparison of the secondary loop outputs to typical RELAP5 outputs as well as a third party source.

  5. MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code

    SciTech Connect

    Dunn, F.E.; Thomas, J.; Liaw, J.; Matos, J.E.

    2008-07-15

    For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO{sub 2} fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)

  6. Analysis of the SL-1 Accident Using RELAPS5-3D

    SciTech Connect

    Francisco, A.D. and Tomlinson, E. T.

    2007-11-08

    On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with a discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).

  7. Waste Evaporator Accident Simulation Using RELAP5 Computer Code

    SciTech Connect

    POLIZZI, L.M.

    2004-04-28

    An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs.

  8. Methodology, status, and plans for development and assessment of the RELAP5 code

    SciTech Connect

    Johnson, G.W.; Riemke, R.A.

    1997-07-01

    RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

  9. Assessment of MIT and UCB wall condensation tests and of the pre-release RELAP5/MOD3.2 code condensation models

    SciTech Connect

    Shumway, R.W.

    1995-01-01

    In recent years, a new class of reactor designs has been proposed that utilize passive safety systems. General Electric has developed a Simplified Boiling Water Reactor (SBWR) design that relies on such passive systems. The SBWR has two passive cooling systems that involve energy transfer by condensation. These are the isolation condenser system (ICS) and the passive containment cooling systems (PCCS). It is important that such heat transfer phenomena be correctly understood and quantified. The General Electric Company has sponsored tests at the Massachusetts Institute of Technology (MIT) and at the University of California at Berkeley (UCB) to obtain data simulating PCCS conditions. Data was obtained with pure steam, steam-air mixtures and steam-helium mixtures. INEL has been contracted by the NRC to evaluate these tests and assess existing condensation heat transfer correlations against the test data. This report assesses the relevance of the tests to SBWR conditions and shows RELAP5/MOD3.2 predictions of the tests.

  10. Thermal Hydraulic Computer Code System.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-16

    Version 00 RELAP5 was developed to describe the behavior of a light water reactor (LWR) subjected to postulated transients such as loss of coolant from large or small pipe breaks, pump failures, etc. RELAP5 calculates fluid conditions such as velocities, pressures, densities, qualities, temperatures; thermal conditions such as surface temperatures, temperature distributions, heat fluxes; pump conditions; trip conditions; reactor power and reactivity from point reactor kinetics; and control system variables. In addition to reactor applications,more » the program can be applied to transient analysis of other thermal‑hydraulic systems with water as the fluid. This package contains RELAP5/MOD1/029 for CDC computers and RELAP5/MOD1/025 for VAX or IBM mainframe computers.« less

  11. Peer review of RELAP5/MOD3 documentation

    SciTech Connect

    Craddick, W.G.

    1993-12-31

    A peer review was performed on a portion of the documentation of the RELAP5/MOD3 computer code. The review was performed in two phases. The first phase was a review of Volume 3, Developmental Assessment problems, and Volume 4, Models and Correlations. The reviewers for this phase were Dr. Peter Griffith, Dr. Yassin Hassan, Dr. Gerald S. Lellouche, Dr. Marino di Marzo and Mr. Mark Wendel. The reviewers recommended a number of improvements, including using a frozen version of the code for assessment guided by a validation plan, better justification for flow regime maps and extension of models beyond their data base. The second phase was a review of Volume 6, Quality Assurance of Numerical Techniques in RELAP5/MOD3. The reviewers for the second phase were Mr. Mark Wendel and Dr. Paul T. Williams. Recommendations included correction of numerous grammatical and typographical errors and better justification for the use of Lax`s Equivalence Theorem.

  12. Recent SCDAP/RELAP5 code applications and improvements

    SciTech Connect

    Harvego, E.A.; Ghan, L.S.; Knudson, D.L.; Siefken, L.J.

    1998-03-01

    This paper summarizes (1) a recent application of the severe accident analysis code SCDAP/RELAP5/MOD3.1, and (2) development and assessment activities associated with the release of SACDAP/RELAP5/MOD3.2. The Nuclear Regulatory Commission (NRC) has been evaluating the integrity of steam generator tubes during severe accidents. MOD3.1 has been used to support that evaluation. Studies indicate that the pressurizer surge line will fail before any steam generator tubes are damaged. Thus, core decay energy would be released as steam through the surge line and the tube wall would be spared from exposure to prolonged flow of high temperature steam. The latest code version, MOD3.2, contains several improvements to models that address both the early phase and late phase of a severe accident. The impact of these improvements to the overall code capabilities has been assessed. Results of the assessment are summarized in this paper.

  13. Analysis of the Peach Bottom Turbine Trip 2 Experiment by Coupled RELAP5-PARCS Three-Dimensional Codes

    SciTech Connect

    Bousbia-Salah, Anis; Vedovi, Juswald; D'Auria, Francesco; Ivanov, Kostadin; Galassi, Giorgio

    2004-10-15

    Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.

  14. Comparison of a RELAP5/MOD2 posttest calculation to the data during the recovery portion of a semiscale single-tube steam generator tube rupture experiment

    SciTech Connect

    Chapman, J.C.

    1986-09-01

    This report discusses the comparisons of a RELAP5 posttest calculation of the recovery portion of the Semiscale Mod-2B test S-SG-1 to the test data. The posttest calculation was performed with the RELAP5/MOD2 cycle 36.02 code without updates. The recovery procedure that was calculated mainly consisted of secondary feed and steam using auxiliary feedwater injection and the atmospheric dump valve of the unaffected steam generator (the steam generator without the tube rupture). A second procedure was initiated after the trends of the secondary feed and steam procedure had been established, and this was to stop the safety injection that had been provided by two trains of both the charging and high pressure injection systems. The Semiscale Mod-2B configuration is a small scale (1/1705), nonnuclear, instrumented, model of a Westinghouse four-loop pressurized water reactor power plant. S-SG-1 was a single-tube, cold-side, steam generator tube rupture experiment. The comparison of the posttest calculation and data included comparing the general trends and the driving mechanisms of the responses, the phenomena, and the individual responses of the main parameters.

  15. Audit calculation of the limiting CESSAR feedwater-line-break transient with RELAP5/MOD1. [PWR

    SciTech Connect

    Chung, K.S.; Kennedy, M.F.; Guttmann, J.

    1983-01-01

    Argonne National Laboratory (ANL) performed a series of audit calculations of the limiting FLB transient presented in Appendix 15B to the CESSAR FSAR, supported by a limited number of additional calculations to investigate the sensitivity of the results (in terms of peak primary reactor system pressure) to break area and reactor trip time. The latter calculations were performed to quantify potential benefits in crediting reactor tip on low steam generator downcomer water level, which occurs earlier than the trip shown in the limiting FSAR transient, which tripped on high pressurizer pressure. These calculations were performed to verify the break spectrum results presented by C-E and to insure that C-E did indeed analyze the limiting transient. All of the ANL calculations were performed with RELAP5/MOD1 (cycle 18) using an input deck developed at ANL from CESSAR plant data provided by C-E. In this paper we compare the results and provide insight into the generic behavior of a Feedwater Line Break transient.

  16. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Yoder, G.L. ); Wendel, M.W. )

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  17. BWR control blade/channel box interaction models for SCDAP/RELAP5

    SciTech Connect

    Griffin, F.P.

    1993-12-01

    The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. Each fuel assembly consists of a fuel rod bundle surrounded by a Zircaloy channel box. Each control blade consists of small stainless steel absorber tubes filled with B{sub 4}C powder surrounded by a stainless steel blade sheath. Under severe accident conditions, material interactions between the B{sub 4}C, stainless steel, and Zircaloy would have a significant impact on the melting and subsequent relocation of the control blade and channel box structures. This paper describes a new BWR control blade/channel box model for the SCDAP/RELAP5 severe accident analysis code that includes the effects of these material interactions. The phenomena represented by this model and the modeling techniques are derived from ORNL analyses of the BWR severe fuel damage experiments. Two examples of the operation of this new model within SCDAP/RELAP5 are provided.

  18. Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

    NASA Astrophysics Data System (ADS)

    Honaiser, Eduardo Henrique Rangel

    The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

  19. RELAP5-3D Architectural Developments in 2004

    SciTech Connect

    Dr. George L. Mesina

    2004-08-01

    Currently, RELAP5 is undergoing a transformation that will replace much of its coding with equivalent structured Fortran 90 coding. Four efforts are underway to modernize the code architecture of RELAP5-3D. These are parallelization, vectorization, code restructuring, and conversion to Fortran 90. The first two improve code run speed via on computer platforms of certain architectures. These code modifications have little effect on normal code performance on non-vector and non-parallel computers because they are mostly done with compiler directives. The third and fourth efforts involve considerable rewriting of the source code. The third code improvement effort addresses code readability and maintainability. These are being greatly enhanced by application of a Fortran code-restructuring tool. The fourth effort is conversion to Fortran 90. The bulk of the coding is being rewritten in Fortran 90. This is a ground up reworking of the coding that begins with completely reorganizing the underlying database and continues with the source code. It will reach every part of RELAP5-3D. Each of these efforts is discussed in detail in a different section. Section 1 relates background information. Section 2 covers the parallelization effort. Section 3 covers the efforts to vectorize the code. Section 4 covers the code restructuring. Section 5 covers the Fortran 90 effort. Outline Background: longevity, maintenance & development, reliability, speed Parallelization: KAI to OpenMP, previous work & current, domain decomposition, done. Vectorization: Speed - Fed init, vectors in PCs, INL Cray SV1, R5 Phant, EXV, results. Code Restructuring: Reason to restructure, study of restruct, For Study: what it does, Fortran 90: Modernization -

  20. SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

    SciTech Connect

    Knudson, Darrell Lee; Ghan, Larry Scott; Dobbe, Charles Albin

    1998-09-01

    Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles, and the steam generator (SG) tubes. The potential for a steam generator tube rupture (SGTR) is of particular concern because fission products could be released to the environment through such a failure. The Nuclear Regulatory Commission (NRC) developed a program to address SG tube integrity issues in operating pressurized water reactors (PWRs) based on the possibility for environmental release. An extensive effort to evaluate the potential for accident-induced SGTRs using SCDAP/RELAP5 at the Idaho National Engineering and Environmental Laboratory (INEEL) was directed as one part of the NRC program. All SCDAP/RELAP5 calculations performed during the INEEL evaluation were based on station blackout accidents (and variations thereof) because those accidents are considered to be one of the more likely scenarios leading to natural circulation flows at temperatures and pressures that could threaten SG tube integrity (as well as the integrity of other vulnerable RCS pressure boundaries). Variations that were addressed included consideration of the effects of RCP seal leaks, intentional RCS depressurization through pressurizer PORVs, SG secondary depressurization, DC-HL bypass flows, U-tube SG sludge accumulation, and quenching of upper plenum stainless steel upon relocation to the lower head. Where available, experimental data was used to guide simulation of natural circulation flows. Independent reviews of the applicability of the natural circulation experimental data, the suitability of the code, and the adequacy of the modeling were completed and review recommendations were incorporated into the evaluation within budget and

  1. Assessment of full power turbine trip start-up test for C. Trillo 1 with RELAP5/MOD2. International Agreement Report

    SciTech Connect

    Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M.

    1993-07-01

    C. Trillo I has developed a model of the plant with RELAP5/MOD2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other PWRs of different design, the main difference being that the reactor is not tripped: a reduction in primary power is carried out instead. Pre-test calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitation Systems, were identified and post-test calculations had been carried out. The results show a good agreement with data for all the compared variables.

  2. Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D

    SciTech Connect

    Cliff B. Davis

    2010-09-01

    RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

  3. A comparison of the RELAP5/MOD3 and PARET/ANL codes with the experimental transient data from the SPERT-IV D-12/25 series.

    SciTech Connect

    Woodruff, W. L.

    1998-01-16

    The results from the RELAP5/MOD3 and PARET/ANL codes are compared with the SPERT-IV series of experimental reactivity insertion transients. The PARET/ANL code provides conservative estimates of SPERT-IV experimental data for the midrange transients and for the more severe transients. The PARET results are similar to the results obtained earlier for the SPERT-I D-12/25 series of experiments. The RELAP5/MOD3 code (including the developmental version 3.2.1.2) gives results comparable to PARET for some midrange transients, but seriously diverges from the experimental data when significant boiling is present. Based on the results of this study, the use of the RELAP5 code for research reactor applications should be limited to transients that do not generate substantial boiling and voids. We hope to be able to resolve these differences in further work with the NRC staff and its contractors. The RELAP5 code would be a more useful tool for the analyses research reactor transients with the addition of suitable correlations for low pressures and plate type geometry.

  4. Application of RELAP5 to a pipe blowdown experiment. [PWR; BWR

    SciTech Connect

    Carlson, K.E.; Ransom, V.H.; Wagner, R.J.

    1980-01-01

    The application of the RELAP5 computer program to a pipe blowdown experiment is described in this paper. The basic hydrodynamic model, constitutive relations, and special process models included in RELAP5 are also briefly discussed. The results of this application confirm the effectiveness of using a choked flow model.

  5. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    SciTech Connect

    Siefken, Larry James; Coryell, Eric Wesley; Paik, Seungho; Kuo, Han Hsiung

    1999-07-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region.

  6. Assessment of RELAP5-3D{copyright} using data from two-dimensional RPI flow tests

    SciTech Connect

    Davis, C.B.

    1998-07-01

    The capability of the RELAP5-3D{copyright} computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code`s logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved.

  7. Hot Zero and Full Power Validation of PHISICS RELAP-5 Coupling

    SciTech Connect

    F. Lodi; C. Rabiti; A. Alfonsi; A. Epiney; M. Sumini

    2013-06-01

    PHISICS is a reactor analysis toolkit developed in the last 3 years at the Idaho National Laboratory that has been also coupled with the thermo-hydraulic plant simulator RELAP5-3D. PHISICS is aimed to provide an optimal trade off between needed computational resources and accuracy in the range of 10~100 cores. In fact this range has been identified as the next 5 to 10 years average computational capability available to nuclear engineer designing and optimizing nuclear reactor cores. Different publication has been already presented [1] showing test of the single modules composing the PHISICS package. Lately the Idaho National Laboratory had the opportunity to access to plant data for the first cycle of a PWR including Hot Zero Power (HZP) and Hot Full Power (HFP). This data provided the opportunity to validate the transport solver, the interpolation capability for mixed macro and micro cross section and the criticality search option of the PHISICS package. In the following we will firstly recall briefly the structure of the different PHISICS modules and then we will illustrate the modeling process and some preliminary results.

  8. International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B. International Agreement Report

    SciTech Connect

    Schultz, R.R.

    1993-12-01

    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.

  9. RELAP5 Model Description and Validation for the BR2 Loss-of-Flow Experiments

    SciTech Connect

    Licht, J. R.; Dionne, B.; Van den Branden, G.; Sikik, E.; Koonen, E.

    2015-07-01

    This paper presents a description of the RELAP5 model, the calibration method used to obtain the minor loss coefficients from the available hydraulic data and the LOFA simulation results compared to the 1963 experimental tests for HEU fuel.

  10. RELAP5 Model of a Two-phase ThermoSyphon Experimental Facility for Fuels and Materials Irradiation

    SciTech Connect

    Carbajo, Juan J; McDuffee, Joel Lee

    2013-01-01

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) does not have a separate materials-irradiation flow loop and requires most materials and all fuel experiments to be placed inside a containment. This is necessary to ensure that internal contaminants such as fission products cannot be released into the primary coolant. As part of the safety basis justification, HFIR also requires that all experiments be able to withstand various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. As with any parallel flow system, HFIR is vulnerable to flow excursion events when vapor is generated in one of those flow paths. The effects of these requirements are to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant and to reduce experiment heat loads to ensure boiling doesn t occur. A new experimental facility for materials irradiation and testing in the HFIR is currently being developed to overcome these limitations. The new facility is unique in that it will have its own internal cooling flow totally independent of the reactor primary coolant and boiling is permitted. The reactor primary coolant will cool the outside of this facility without contacting the materials inside. The ThermoSyphon Test Loop (TSTL), a full scale prototype of the proposed irradiation facility to be tested outside the reactor, is being designed and fabricated (Ref. 1). The TSTL is a closed system working as a two-phase thermosyphon. A schematic is shown in Fig. 1. The bottom central part is the boiler/evaporator and contains three electric heaters. The vapor generated by the heaters will rise and be condensed in the upper condenser, the condensate will drain down the side walls and be circulated via a downcomer back into the bottom of the boiler. An external flow system provides coolant that simulates the HFIR primary coolant

  11. A station blackout simulation for the Advanced Neutron Source Reactor using the integrated primary and secondary system model

    SciTech Connect

    Schneider, E.A.

    1994-06-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at Oak Ridge National Laboratory. This paper deals with thermal-hydraulic analysis of ANSR`s cooling systems during nominal and transient conditions, with the major effort focusing upon the construction and testing of computer models of the reactor`s primary, secondary and reflector vessel cooling systems. The code RELAP5 was used to simulate transients, such as loss of coolant accidents and loss of off-site power, as well as to model the behavior of the reactor in steady state. Three stages are involved in constructing and using a RELAP5 model: (1) construction and encoding of the desired model, (2) testing and adjustment of the model until a satisfactory steady state is achieved, and (3) running actual transients using the steady-state results obtained earlier as initial conditions. By use of the ANSR design specifications, a model of the reactor`s primary and secondary cooling systems has been constructed to run a transient simulating a loss of off-site power. This incident assumes a pump coastdown in both the primary and secondary loops. The results determine whether the reactor can survive the transition from forced convection to natural circulation.

  12. Assessment of RELAP5/MOD3 with the LOFT L9-1/L3-3 experiment simulating an anticipated transient with multiple failures

    SciTech Connect

    Bang, Y.S.; Seul, K.W.; Kim, H.J.

    1994-02-01

    The RELAP5/MOD3 5m5 code is assessed using the L9-1/L3-3 test carried out in the LOFT facility, a 1/60-scaled experimental reactor, simulating a loss of feedwater accident with multiple failures and the sequentially-induced small break loss-of-coolant accident. The code predictability is evaluated for the four separated sub-periods with respect to the system response; initial heatup phase, spray and power operated relief valve (PORV) cycling phase, blowdown phase and recovery phase. Based on the comparisons of the results from the calculation with the experiment data, it is shown that the overall thermal-hydraulic behavior important to the scenario such as a heat removal between the primary side and the secondary side and a system depressurization can be well-predicted and that the code could be applied to the full-scale nuclear power plant for an anticipated transient with multiple failures within a reasonable accuracy. The minor discrepancies between the prediction and the experiment are identified in reactor scram time, post-scram behavior in the initial heatup phase, excessive heatup rate in the cycling phase, insufficient energy convected out the PORV under the hot leg stratified condition in the saturated blowdown phase and void distribution in secondary side in the recovery phase. This may come from the code uncertainties in predicting the spray mass flow rate, the associated condensation in pressurizer and junction fluid density under stratified condition.

  13. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  14. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    SciTech Connect

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-31

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  15. Implementation of DOWTHERM A Properties into RELAP5-3D/ATHENA

    SciTech Connect

    Richard L. Moore

    2010-04-01

    DOWTHERM A oil is being considered for use as a heat transfer fluid in experiments to help in the design of heat transfer components for the Next Generation Nuclear Plant (NGNP). In conjection with the experiments RELAP5-3D/ATHENA will be used to help design and analyzed the data generated by the experiments. Inorder to use RELAP5-3D the thermophysical properties of DOWTHERM A were implemented into the fluids package of the RELAP5-3D/ATHENA computer propgram. DOWTHERM A properties were implemented in RELAP5-3D/ATHENA using thermophysical property data obtain from a Dow Chemical Company brochure. The data were curve fit and the polynomial equations developed for each required property were input into a fluid property generator. The generated data was then compared to the orginal DOWTHERM A data to verify that the fluid property data generated by the RELAP5-3D/ATHENA code was representitive of the original input data to the generator.

  16. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  17. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

    SciTech Connect

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

  18. Development and New Directions for the RELAP5-3D Graphical Users Interface

    SciTech Connect

    Mesina, George Lee

    2001-09-01

    The direction of development for the RELAP5 Graphical User Interfaces (RGUI) has been extended. In addition to existing plans for displaying all aspects of RELAP5 calculations, the plan now includes plans to display the calculations of a variety of codes including SCDAP, RETRAN and FLUENT. Recent work has included such extensions along with the previously planned and user-requested improvements and extensions. Visualization of heat-structures has been added. Adaptations were made for another computer program, SCDAP-3D, including plant core views. An input model builder for generating RELAP5-3D input files was partially implemented. All these are reported. Plans for future work are also summarized. These include an input processor that transfers steady-state conditions into an input file.

  19. Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code

    NASA Astrophysics Data System (ADS)

    De Rosa, F.; Lombardo, C.; Mascari, F.; Polidori, M.; Chiovaro, P.; D'Amico, S.; Moscato, I.; Vella, G.

    2014-11-01

    The present paper deals with the investigation of the evolution and consequences of a Station Black-Out (SBO) initiating event transient in the SPES3 facility [1]. This facility is an integral simulator of a small modular reactor being built at the SIET laboratories, in the framework of the R&D program on nuclear fission funded by the Italian Ministry of Economic Development and led by ENEA. The SBO transient will be simulated by using the RELAP5 and TRACE nodalizations of the SPES3 facility. Moreover, the analysis will contribute to study the differences on the code predictions considering the different modelling approach with one and/or three-dimensional components and to compare the capability of these codes to describe the SPES3 facility behaviour.

  20. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    SciTech Connect

    Pecchia, M.; D'Auria, F.; Mazzantini, O.

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  1. Assessment of RELAP5/MOD2 against a turbine trip from 100% power in the Vandellos II nuclear power plant

    SciTech Connect

    Llopis, C. ); Perez, J.; Mendizabal, R. )

    1993-06-01

    An assessment of RELAP5/MOD2 cycle 36.04 against a turbine trip from 100% power in the Vandellos II NPP (Spain) is presented. The work is inscribed in the framework of the Spanish contribution to ICAP Project. The model used in the simulation consists of a single loop, a steam generator and a steam line up to the steam header all of them enlarged on a scale of 3:1; and full-scaled reactor vessel and pressurizer. The results of calculations have been in reasonable agreement with plant measurements. An additional study has been performed to check the ability of a model in which all the plant components are full-scaled to reproduce the transient. A second study has been performed using the Homogeneous Equilibrium Model in the pressurizer, trying to elucidate the influence of the velocity slip in the primary depressurization rate.

  2. A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3

    SciTech Connect

    Andreani, M.; Analytis, G.T.; Aksan, S.N.

    1995-09-01

    The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.

  3. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    SciTech Connect

    Wissinger, G.; Klingenfus, J.

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  4. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    SciTech Connect

    E. W. Coryell; L. J. Siefken; S. Paik

    1998-09-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region.

  5. Transient prediction of 19-tube once-through steam generator by RELAP5/MOD1

    SciTech Connect

    Hassan, Y.A.; Morgan, C.D.

    1982-01-01

    A simulation of Babcock and Wilcox's Alliance Research Center loss-of-feedwater of 19-tube model of once-through steam generator (OTSG) was performed with RELAP5/MOD1 and compared with the experimental data. Acceptable transient scenario was obtained when implementing Biasi and Macbeth critical heat flux correlations.

  6. Recent Hydrodynamics Improvements to the RELAP5-3D Code

    SciTech Connect

    Richard A. Riemke; Cliff B. Davis; Richard.R. Schultz

    2009-07-01

    The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model.

  7. SCDAP/RELAP5 modeling of heat transfer and flow losses in lower head porous debris. Revision 1

    SciTech Connect

    Siefken, L.J.; Coryell, E.W.; Paik, S.; Kuo, H.

    1999-05-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate ma nner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region.

  8. Implementation of Molten Salt Properties into RELAP5-3D/ATHENA

    SciTech Connect

    Cliff Davis

    2005-01-01

    Molten salts are being considered as coolants for the Next Generation Nuclear Plant (NGNP) in both the reactor and the heat transport loop between the reactor and the hydrogen production plant because of their superior thermophysical properties compared to helium. Because specific molten salts have not been selected for either application, four separate molten salts were implemented into the RELAP5-3D/ATHENA computer program as working fluids. The implemented salts were LiF-BeF2 in a molar mixture that is 66% LiF and 34% BeF2, respectively, NaBF4-NaF (92% and 8%), LiF-NaF-KF (11.5%, 46.5%, and 42%), and NaF-ZrF4 (50% and 50%). LiF-BeF2 is currently the first choice for the primary coolant for the Advanced High- Temperature Reactor, while NaF-ZrF4 is being considered as an alternate. NaBF4-NaF and LiFNaF- KF are being considered as possible coolants for the heat transport loop. The molten salts were implemented into ATHENA using a simplified equation of state based on data and correlations obtained from Oak Ridge National Laboratory. The simplified equation of state assumes that the liquid density is a function of temperature and pressure and that the liquid heat capacity is constant. The vapor is assumed to have the same composition as the liquid and is assumed to be a perfect gas. The implementation of the thermodynamic properties into ATHENA for LiF-BeF2 was verified by comparisons with results from a detailed equation of state that utilized a soft-sphere model. The comparisons between the simplified and soft-sphere models were in reasonable agreement for liquid. The agreement for vapor properties was not nearly as good as that obtained for liquid. Large uncertainties are possible in the vapor properties because of a lack of experimental data. The simplified model used here is not expected to be accurate for boiling or single-phase vapor conditions. Because neither condition is expected during NGNP applications, the simplified equation of state is considered

  9. Assessment of RELAP5/MOD2 against a load rejection from 100% to 50% power in the Vandellos II nuclear power plant. International Agreeement Report

    SciTech Connect

    Llopis, C.; Mendizabal, R.; Perez, J.

    1993-06-01

    An assessment of RELAP5/MOD2 cycle 36.04 against a load rejection from 100% to 50% power in Vandals II NPP (Spain) is presented. The work is inscribed in the framework of the Spanish contribution to ICAP Project. The model used in the simulation consists of a single loop, a steam generator and a steam line up to the steam header all of them enlarged on a scale of 3:1, and full-scaled reactor vessel and pressurizer. The results of the calculations have been in reasonable agreement with plant measurements.

  10. Assessment of RELAP5/MOD2 against a load rejection from 100% to 50% power in the Vandellos II nuclear power plant

    SciTech Connect

    Llopis, C. ); Mendizabal, R.; Perez, J. )

    1993-06-01

    An assessment of RELAP5/MOD2 cycle 36.04 against a load rejection from 100% to 50% power in Vandals II NPP (Spain) is presented. The work is inscribed in the framework of the Spanish contribution to ICAP Project. The model used in the simulation consists of a single loop, a steam generator and a steam line up to the steam header all of them enlarged on a scale of 3:1, and full-scaled reactor vessel and pressurizer. The results of the calculations have been in reasonable agreement with plant measurements.

  11. Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2. International Agreement Report

    SciTech Connect

    Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.

    1993-12-01

    The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ``Pressurizer spray valve faulty opening`` presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data.

  12. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  13. SCDAP/RELAP5/MOD2 code manual

    SciTech Connect

    Hohorst, J.K. )

    1990-02-01

    This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.

  14. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R. ); Choi, C.J. )

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  15. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  16. RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

    SciTech Connect

    Lombardi Costa, Antonella; Petruzzi, Alessandro; D'Auria, Francesco

    2006-07-01

    Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)

  17. Heat Transfer Boundary Conditions in the RELAP5-3D Code

    SciTech Connect

    Richard A. Riemke; Cliff B. Davis; Richard R. Schultz

    2008-05-01

    The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.

  18. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    SciTech Connect

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  19. RELAP5-3D Modeling of Heat Transfer Components (Intermediate Heat Exchanger and Helical-Coil Steam Generator) for NGNP Application

    SciTech Connect

    N. A. Anderson; P. Sabharwall

    2014-01-01

    The Next Generation Nuclear Plant project is aimed at the research and development of a helium-cooled high-temperature gas reactor that could generate both electricity and process heat for the production of hydrogen. The heat from the high-temperature primary loop must be transferred via an intermediate heat exchanger to a secondary loop. Using RELAP5-3D, a model was developed for two of the heat exchanger options a printed-circuit heat exchanger and a helical-coil steam generator. The RELAP5-3D models were used to simulate an exponential decrease in pressure over a 20 second period. The results of this loss of coolant analysis indicate that heat is initially transferred from the primary loop to the secondary loop, but after the decrease in pressure in the primary loop the heat is transferred from the secondary loop to the primary loop. A high-temperature gas reactor model should be developed and connected to the heat transfer component to simulate other transients.

  20. PARET/ANL and RELAP5/MOD2 benchmarking comparison with the Spert-IV test data

    SciTech Connect

    Kim, S.S.; McKibben, J.C.

    1989-01-01

    Results of PARET/ANL and RELAP5/MOD2 computations on one of the Spert-IV tests are compared to select the code that best predicts the peak power and fuel plate temperature resulting from reactivity-induced transients for use in the University of Missouri Research Reactor (MURR) upgrade safety-related analysis. The D-12/25 core of the Spert-IV tests was selected for comparison because the test was performed under forced coolant circulation in a low-pressure and low-temperature environment, and this test used plate-type fuel as does MURR. The square-shaped D-12/25 core consisted of a 5 {times} 5 array of 20 fuel assemblies, 4 control rod assemblies, and 1 transient rod assembly. Control of the reactor was accomplished by the use of four boron/aluminum control rods, and the power excursion was initiated by a step reactivity addition established by ejecting the poison section of the transient rod from the core.

  1. Import Manipulate Plot RELAP5/MOD3 Data

    SciTech Connect

    Jones, K. R.

    1999-10-05

    XMGR5 was derived from an XY plotting tool called ACE/gr, which is copyrighted by Paul J. Turner and in the public domain. The interactive version of ACE/GR is xmgr, and includes a graphical interface to the X-windows system. Enhancements to xmgr have been developed which import, manipualate, and plot data from RELAP/MOD3, MELCOR, FRAPCON, and SINDA codes, and NRC databank files. capabilities, include two-phase property table lookup functions, an equation interpreter, arithmetic library functions, and units conversion. Plot titles, labels, legends, and narrative can be displayed using Latin or Cyrillic alphabets.

  2. Import Manipulate Plot RELAP5/MOD3 Data

    Energy Science and Technology Software Center (ESTSC)

    1999-10-05

    XMGR5 was derived from an XY plotting tool called ACE/gr, which is copyrighted by Paul J. Turner and in the public domain. The interactive version of ACE/GR is xmgr, and includes a graphical interface to the X-windows system. Enhancements to xmgr have been developed which import, manipualate, and plot data from RELAP/MOD3, MELCOR, FRAPCON, and SINDA codes, and NRC databank files. capabilities, include two-phase property table lookup functions, an equation interpreter, arithmetic library functions, andmore » units conversion. Plot titles, labels, legends, and narrative can be displayed using Latin or Cyrillic alphabets.« less

  3. Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

    SciTech Connect

    Low, J.O.; Schmitt, B.E.

    1988-02-01

    A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may be exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.

  4. Savannah River Site reactor hardware design modification study

    SciTech Connect

    Fisher, J.E.

    1990-01-01

    A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs.

  5. Reactor System Transient Code.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  6. Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B W Plant: An application of the CSAU methodology using the RELAP5/MOD3 computer code

    SciTech Connect

    Ortiz, M G; Ghan, L S

    1992-12-01

    The Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability, and Uncertainty (CSAU) to evaluate best estimate code uncertainties. The objective of this work was to adapt and demonstrate the CSAU methodology for a small-break loss-of-coolant accident (SBLOCA) in a Pressurized Water Reactor of Babcock Wilcox Company lowered loop design using RELAP5/MOD3 as the simulation tool. The CSAU methodology was successfully demonstrated for the new set of variants defined in this project (scenario, plant design, code). However, the robustness of the reactor design to this SBLOCA scenario limits the applicability of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses.

  7. Improvements to the RELAP5-3D Nearly-Implicit Numerical Scheme

    SciTech Connect

    Richard A. Riemke; Walter L. Weaver; RIchard R. Schultz

    2005-05-01

    The RELAP5-3D computer program has been improved with regard to its nearly-implicit numerical scheme for twophase flow and single-phase flow. Changes were made to the nearly-implicit numerical scheme finite difference momentum equations as follows: (1) added the velocity flip-flop mass/energy error mitigation logic, (2) added the modified Henry-Fauske choking model, (3) used the new time void fraction in the horizontal stratification force terms and gravity head, and (4) used an implicit form of the artificial viscosity. The code modifications allow the nearly-implicit numerical scheme to be more implicit and lead to enhanced numerical stability.

  8. Developmental assessment of the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    SciTech Connect

    Hanson, R.G.; Johnson, E.C.; Carlson, K.E.; Chou, C.Y.; Davis, C.B.; Martin, R.P.; Riemke, R.A.; Wagner, R.J.

    1992-07-01

    This report documents ten developmental assessment problems which were used to test the multidimensional component in RELAP5/MOD2.5, Version 3w. The problems chosen were a rigid body rotation problem, a pure radial symmetric flow problem, an r-[theta] symmetric flow problem, a fall problem, a rest problem, a basic one-dimensional flow test problem, a gravity wave problem, a tank draining problem, a flow through the center problem, and coverage analysis using PIXIE. The multidimensional code calculations are compared to analytical solutions and one-dimensional code calculations. The discussion section of each problem contains information relative to the code's ability to simulate these problems.

  9. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  10. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  11. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect

    Cliff B. Davis

    2007-09-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  12. Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

    SciTech Connect

    Hohorst, J.K.; Allison, C.M.

    1992-12-31

    The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons.

  13. Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

    SciTech Connect

    Hohorst, J.K.; Allison, C.M.

    1992-01-01

    The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons.

  14. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  15. Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU (Korea Nuclear Unit) No. 1 Plant

    SciTech Connect

    Chung, Bud-Dong; Kim, Hho-Jung . Korea Nuclear Safety Center); Lee, Young-Jin )

    1990-04-01

    This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU {number sign}1 (Korea Nuclear Unit Number 1). KNU {number sign}1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs.

  16. Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

    NASA Astrophysics Data System (ADS)

    Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

    2012-10-01

    Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

  17. RELAP5-3D Developmental Assessment. Comparison of Version 4.3.4i on Linux and Windows

    SciTech Connect

    Bayless, Paul David

    2015-10-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.3i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

  18. RELAP5-3D Developmental Assessment: Comparison of Version 4.2.1i on Linux and Windows

    SciTech Connect

    Paul D. Bayless

    2014-06-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.2i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

  19. Independent code assessment at BNL in FY 1982. [TRAC-PF1; RELAP5/MOD1; TRAC-BD1

    SciTech Connect

    Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G.; Yuelys-Miksis, C.

    1982-01-01

    Independent assessment of the advanced codes such as TRAC and RELAP5 has continued at BNL through the Fiscal Year 1982. The simulation tests can be grouped into the following five categories: critical flow, counter-current flow limiting (CCFL) or flooding, level swell, steam generator thermal performance, and natural circulation. TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes were assessed by simulating all of the above experiments, whereas the TRAC-BD1 (Version 12.0) code was applied only to the CCFL tests. Results and conclusions of the BNL code assessment activity of FY 1982 are summarized below.

  20. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  1. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  2. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  3. ICAP (International Code Assessment and Applications Program) assessment of RELAP5/MOD2, Cycle 36. 05 against LOFT (Loss of Fluid Test) Small Break Experiment L3-7

    SciTech Connect

    Lee, Euy-Joon; Chung, Bud-Dong; Kim, Hho-Jung . Korea Nuclear Safety Center)

    1990-04-01

    The LOFT small break (1 in-dia) experiment L3-7 has been analyzed using the reactor thermal hydraulic analysis code RELAP5/MOD2, Cycle 36.05. The base calculation (Case A) was completed and compared with the experimental data. Three types of sensitivity studies (Cases B, Cm, and D) were carried out to investigate the effects of (1) break discharge coefficient Cd, (2) pump two-phase difference multiplier and (3) High Pressure Injection System (HPIS) capacity on major thermal and hydraulic (T/H) parameters. A nodalization study (Case E) was conducted to assess the phenomena with a simplified nodalization. The results indicate that Cd of 0.9 and 0.1 fit to the single discharge flow rate of Test L3-7 best among the tried cases. The pump two-phase multiplier has little effects on the T/H parameters because of the low discharge flow rate and the early pump coast down in this smaller size SBLOCA. But HPIS capacity has a very strong influence on parameters such as pressure, flow and temperature. It is also shown that a simplified nodalization could accomodate the dominant T/H phenomena with the same degree of code accuracy and efficiency.

  4. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  5. Developmental assessment of the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    SciTech Connect

    Hanson, R.G.; Johnson, E.C.; Carlson, K.E.; Chou, C.Y.; Davis, C.B.; Martin, R.P.; Riemke, R.A.; Wagner, R.J.

    1992-07-01

    This report documents ten developmental assessment problems which were used to test the multidimensional component in RELAP5/MOD2.5, Version 3w. The problems chosen were a rigid body rotation problem, a pure radial symmetric flow problem, an r-{theta} symmetric flow problem, a fall problem, a rest problem, a basic one-dimensional flow test problem, a gravity wave problem, a tank draining problem, a flow through the center problem, and coverage analysis using PIXIE. The multidimensional code calculations are compared to analytical solutions and one-dimensional code calculations. The discussion section of each problem contains information relative to the code`s ability to simulate these problems.

  6. Evaluation and assessment of reflooding models in RELAP5/Mod2.5 and RELAP5/Mod3 codes using Lehigh University and PSI-Neptun bundle experimental data

    SciTech Connect

    Sencar, M.; Aksan, N.

    1995-09-01

    An extensive analysis and assessment work on reflooding models of RELAP5/Mod2.5 and, RELAP5/Mod3/v5m5 and RELAP/Mod3/v7j have been performed. Experimental data from LehighUniversityv. and PSI-NEPTUN bundle reflooding experiments have been used for the assessment, since both of these tests cover a broad range of initial conditions. Within the range of these initial conditions, it was tried to identify their separate impacts on the calculated results. A total of six Lehigh University reflooding bundle tests and two PSI-NEPTUN tests with bounding initial conditions are selected for the analysis. Detailed nodalisation studies both for hydraulic and conduction heat transfer were done. On the basis of the results obtained from these cases, a base nodalisation scheme was established. All the other analysis work was performed by using this base nodalisation. RELAP5/Mod2.5 results do not change with renodalisation but RELAP5/Mod3 results are more sensitive to renodalisation. The results of RELAP5/Mod2.5 versions show very large deviations from the used experimental data. These results indicate that some of the phenomenology of the events occurring during the reflooding could not be identified. In the paper, detailed discussions on the main reasons of the deviations from the experimental data will be presented. Since, the results and findings of this study are meant to be a developmental aid, some recommendations have been drawn and some of these have already been implemented at PSI with promising results.

  7. Modeling a Printed Circuit Heat Exchanger with RELAP5-3D for the Next Generation Nuclear Plant

    SciTech Connect

    Not Available

    2010-12-01

    The main purpose of this report is to design a printed circuit heat exchanger (PCHE) for the Next Generation Nuclear Plant and carry out Loss of Coolant Accident (LOCA) simulation using RELAP5-3D. Helium was chosen as the coolant in the primary and secondary sides of the heat exchanger. The design of PCHE is critical for the LOCA simulations. For purposes of simplicity, a straight channel configuration was assumed. A parallel intermediate heat exchanger configuration was assumed for the RELAP5 model design. The RELAP5 modeling also required the semicircular channels in the heat exchanger to be mapped to rectangular channels. The initial RELAP5 run outputs steady state conditions which were then compared to the heat exchanger performance theory to ensure accurate design is being simulated. An exponential loss of pressure transient was simulated. This LOCA describes a loss of coolant pressure in the primary side over a 20 second time period. The results for the simulation indicate that heat is initially transferred from the primary loop to the secondary loop, but after the loss of pressure occurs, heat transfers from the secondary loop to the primary loop.

  8. Theory and input requirements for the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

    SciTech Connect

    Hanson, R.G.; Johnson, E.C.; Carlson, K.E.; Riemke, R.A.; Wagner, R.J.

    1992-07-01

    This report documents the theory and input requirements for the multidimensional component in RELAP5/MOD2.5, Version 3w. The equations in Cartesian and cylindrical coordinates are presented as well as the shallow water terms. The implementation of these equations is then discussed. Finally, the constitutive models and input requirements are then described.

  9. Reactor safety assessment system

    SciTech Connect

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  10. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  11. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  12. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  13. RELAP5/MOD3 assessment using the Semiscale 50% Feed Line Break test S-FS-11

    SciTech Connect

    Lee, E.J.; Chung, B.D.; Kim, H.J.

    1993-06-01

    The RELAP5/MOD3 5m5 code was assessed using the 1/1705 volume scaled Semiscale 50% Feed Line Break (FLB) test S-FS-11. Test S-FS-11 was designed in three phases: (a) blowdown phase, (b) stabilization phase, and (c) refill phase. The first objective was to assess the code applicability to 50% FLB situation, the second was to evaluate the FSAR conservatisms regarding SG heat transfer degradation, steam line check valve failure, break flow state, and peak primary system pressure, and the third was to validate the EOP effectiveness. The code was able to simulate the major T/H parameters except for the two-phase break flow and the secondary convective heat transfer rate. The two-phase break flow had still deficiencies. The current boiling heat transfer rate was developed from the data for flow inside of a heated tube, not for flow around heated tubes in a tube bundle. Results indicated that the assumption of 100% heat transfer until the liquid inventory depletion was not conservative, the failed affected steam generator main steam line check valve assumption was not either conservative, the measured break flow experienced all types of flow conditions, the relative proximity to the 110% design pressure limit was conservative. The automatic actions during the blowdown phase were effective in mitigating the consequences. The stabilization operation performed by operator actions were effective to permit natural circulation cooldown and depressurization. The voided secondary refill operations also verified the effectiveness of the operations while recovering the inventory in a voided steam generator.

  14. Assessment of the reflood oxidation models in SCDAP/RELAP5/MOD3.1

    SciTech Connect

    Hohorst, J.K.; Allison, C.M.

    1996-04-01

    Reflooding of a hot damaged core following the start of a severe accident can lead to significant increases in the heating, melting, and oxidation of the core prior to the termination of the accident. These effects have been observed in bundle heating and melting experiments terminated by the addition of water and are postulated to have had a major impact on the accident progression in the TMI-2 accident. Although the detailed mechanisms for the processes are not completely understood, new SCDAP/RELAP5/MOD3.1e models, describing the cracking /spalling of oxidized fuel rod cladding during reflood, and the resulting oxidation of the underlying Zircaloy and relocating liquefied U-Zr-O, provide a reasonable estimate of the experimentally-observed bundle temperatures, hydrogen production, and changes in bundle geometry. This paper provides a brief description of the new models, selected highlights from code-to-data comparisons, and selected results from a recent set of calculations fro TMI-2 using the new models. The potential impact of these new models on other plant calculations is discussed in the concluding portion of this paper.

  15. Independent assessment of TRAC and RELAP5 codes through separate effects tests

    SciTech Connect

    Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G.; Yuelys-Miksis, C.; Pu, J.

    1983-01-01

    Independent assessment of TRAC-PF1 (Version 7.0), TRAC-BD1 (Version 12.0) and RELAP5/MOD1 (Cycle 14) that was initiated at BNL in FY 1982, has been completed in FY 1983. As in the previous years, emphasis at Brookhaven has been in simulating various separate-effects tests with these advanced codes and identifying the areas where further thermal-hydraulic modeling improvements are needed. The following six catetories of tests were simulated with the above codes: (1) critical flow tests (Moby-Dick nitrogen-water, BNL flashing flow, Marviken Test 24); (2) Counter-Current Flow Limiting (CCFL) tests (University of Houston, Dartmouth College single and parallel tube test); (3) level swell tests (G.E. large vessel test); (4) steam generator tests (B and W 19-tube model S.G. tests, FLECHT-SEASET U-tube S.G. tests); (5) natural circulation tests (FRIGG loop tests); and (6) post-CHF tests (Oak Ridge steady-state test).

  16. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    SciTech Connect

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  17. Reformulation RELAP5-3D in FORTRAN 95 and Results

    SciTech Connect

    Dr. George L Mesina

    2010-08-01

    RELAP5-3D is a nuclear power plant code used worldwide for safety analysis, design, and operator training. In keeping with ongoing developments in the computing industry, we have re-architected the code in the FORTRAN 95 language, the current, fully-available, FORTRAN language. These changes include a complete reworking of the database and conversion of the source code to take advantage of new constructs. The improvements and impacts to the code are manifold. It is a completely machine-independent code that produces machine independent fluid property and plot files and expands to the exact size needed to accommodate the user’s input. Runtime is generally better for larger input models. Other impacts of code conversion are improved code readability, reduced maintenance and development time, increased adaptability to new computing platforms, and increased code longevity. The conversion methodology, code improvements and testing upgrades are presented in a manner that will be useful to future conversion projects for other such large codes. Comparison between the pre- and post-conversion code are made on the basis of code metrics and code performance.

  18. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-09-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios.

  19. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  20. The influence of CHF prediction on modeling of once-through steam generators using RELAP5/MOD1

    SciTech Connect

    Hassan, Y.A.; Morgan, C.D.

    1982-01-01

    Comparisons of the predictions of RELAP5/MOD1 to data obtained from a 19-tube model of a once-through steam generator (OTSG) were performed. The initial results were not satisfactory since the predicted outlet steam temperature was much too low. This discrepancy was traced to the inappropriate use of the modified Zuber critical heat flux (CHF) correlation for the conditions occurring during integral economizer OTSG operation. A study of available low-flow CHF correlations was performed which showed that either the Macbeth or Biasi correlations used in conjunction with RELAP5/MOD1 would produce good agreement with the data for the integral economizer type OTSG. The Macbeth correlation was the best for the OTSG with a recirculation path; however, it was not entirely satisfactory due to a slight delay in its prediction of CHF.

  1. Steady-state and transient prediction of a 19-tube once-through steam generator using RELAP5/MOD1

    SciTech Connect

    Hassan, Y.A.; Morgan, C.D.

    1983-01-01

    Comparisons of the predictions of RELAP5/MOD1 to data obtained from a 19-tube model of a once-through steam generator (OTSG) were performed. The initial results were not satisfactory since the predicted outlet steam temperature was much too low. This discrepancy was traced to the inappropriate use of the modified Zuber critical heat flux (CHF) correlation for the conditions occurring during integral economizer OTSG operation. A study of available low-flow CHF correlations was performed that showed that either the Macbeth or Biasi correlations used in conjunction with RELAP5/MOD1 would produce good agreement with both the steadystate and transient data for the integral economizertype OTSG. The Macbeth correlation was the best for the OTSG with a recirculation path; however, it was not entirely satisfactory due to a slight delay in its prediction of CHF. A loss-of-feedwater transient was modeled using the Macbeth CHF correlation and compared to experimental data with satisfactory results.

  2. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    SciTech Connect

    Not Available

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  3. RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2

    SciTech Connect

    Perez, J.; Mendizabal, R.

    1992-04-01

    This document presents the analysis of the OECD LOFT LP-SB-2 Experiment performed by the Consejo de Seguridad Nuclear of Spain working group making use of RELAP5/MOD2 in the frame of the Spanish LOFT Project. LB-SB-2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3-inch equivalent diameter break in the hot leg of a commercial PWR.

  4. Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

    SciTech Connect

    Bovalini, R.; D`Auria, F.; Galassi, G.M.; Mazzini, M.

    1996-11-01

    RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool of ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions.

  5. Assessment of core damage models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

    SciTech Connect

    Coryell, E.W.

    1991-12-31

    The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel.

  6. Assessment of core damadge models in SCDAP/RELAP5 during OECD LOFT LP-FP-2

    SciTech Connect

    Coryell, E.W.

    1991-01-01

    The US Nuclear Regulatory Commission has sponsored a program to apply the SCDAP/RELAP5 code to analysis of the transient and reflood phases of the OECD LOFT LP-FP-2 Experiment. The principal objectives of the LP-FP-2 experiment were to determine the fission product release from the fuel during the early phases of a severe fuel damage scenario and to examine the phenomena controlling fission product transport in a vapor/aerosol environment. Calculations with the SCDAP/RELAP5 code, developed at the INEL with NRC support, have been performed to (1) examine the phenomena controlling the progression of both transient and reflood phases of the experiment, (2) enhance our understanding of the phenomena occurring during reflood and add credence to the postulated phenomenological sequence, (3) assess the ability of SCDAP/RELAP5 to examine severe fuel damage issues and phenomena, and (4) identify code strengths and deficiencies with the intent of prioritizing code improvements. Results indicate that the code is able to analyze the early phases of severe fuel damage reasonably well, with potential deficiencies in modelling interaction between molten control rod material and intact fuel.

  7. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. )

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.

  8. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  9. SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head

    SciTech Connect

    Siefken, Larry James; Harvego, Edwin Allan

    2000-04-01

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head.

  10. SCDAP/RELAP5 modeling of movement of melted material through porous debris in lower head

    SciTech Connect

    L. J. Siefken; E. A. Harvego

    2000-04-02

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material may permeate to near the bottom of a 1m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material was calculated to cause a 12% increase in the heat flux on the external surface of the lower head.

  11. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  12. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  13. Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests. International Agreement Report

    SciTech Connect

    Cho, S.; Arne, N.; Chung, B.D.; Kim, H.J.

    1993-06-01

    The CCFL model used in RELAP5/MOD3 version 5m5 has been assessed against simple vertical tubes and bundle tests performed at a facility of Korea Atomic Energy Research Institute. The effect of changes in tube diameter and nodalization of tube section were investigated. The roles of interfacial drags on the flooding characteristics are discussed. Differences between the calculation and the experiment are also discussed. A comparison between model assessment results and the test data showed that the calculated value lay well on the experimental flooding curve specified by user, but the pressure jump before onset of flooding was not calculated.

  14. RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

    SciTech Connect

    Moore, R.L.; Sloan, S.M.; Schultz, R.R.; Wilson, G.E.

    1996-10-01

    Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available.

  15. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  16. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  17. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  18. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  19. RELAP5-3D Transient Modelling for NGNP Integrated Plant

    SciTech Connect

    Sabharwall, P.; Anderson, N. A.

    2014-06-01

    The High-Temperature Gas-cooled Reactor (HTGR) is designed with outlet temperatures ranging between 750°C and 800°C. These high outlet temperatures enhance the power production efficiency and facilitate a variety of industrial applications. The objective of this study is to understand the response of the primary system to potential transients in the secondary system. For this analysis, the transient condition originates in the Intermediate Heat Exchanger (IHX) or Steam Generator (SG) of the HTGR-integrated plant. The transients analysed are: a loss of pressure; loss of feedwater flow; inadvertent closure of main steam valve; decrease in returning gas temperature and heat load step change. The results show a large dependence on the negative reactivity added to the fuel as a function of increased temperature. The returning gas temperature decrease transient resulted in the highest fuel temperature (1361°C). Fuel temperature was shown to be less than the 1600°C fuel limit for each case analysed.

  20. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  1. Assessment of RELAP5/MOD2 against a pressurizer spray valve inadverted fully opening transient and recovery by natural circulation in Jose Cabrera Nuclear Station

    SciTech Connect

    Arroyo, R.; Rebollo, L.

    1993-06-01

    This document presents the comparison between the simulation results and the plant measurements of a real event that took place in JOSE CABRERA nuclear power plant in August 30th, 1984. The event was originated by the total, continuous and inadverted opening of the pressurizer spray valve PCV-400A. JOSE CABRERA power plant is a single loop Westinghouse PWR belonging to UNION ELECTRICA FENOSA, S.A. (UNION FENOSA), an Spanish utility which participates in the International Code Assessment and Applications Program (ICAP) as a member of UNIDAD ELECTRICA, S.A. (UNESA). This is the second of its two contributions to the Program: the first one was an application case and this is an assessment one. The simulation has been performed using the RELAP5/MOD2 cycle 36.04 code, running on a CDC CYBER 180/830 computer under NOS 2.5 operating system. The main phenomena have been calculated correctly and some conclusions about the 3D characteristics of the condensation due to the spray and its simulation with a 1D tool have been got.

  2. Assessment of TRAC-PF1 and RELAP5/MOD1 codes with GE large-vessel blowdown test

    NASA Astrophysics Data System (ADS)

    Jo, J. H.

    1983-06-01

    The large vessel blowdown Test No. 5801-15 was simulated with the TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes. The test facility consisted of a pressure vessel, 49 in. in diameter by 14 ft long, a 2.5 in. diameter converging-diverging nozzle and a blowdown line connected to the center of the upper part of the vessel (elevation from the bottom of the vessel 10.5 ft). The vessel was filled with saturated water up to 5.5 ft at 1060 psia. The test was initiated by rupturing a disc attached at the end of the nozzle. Blowdown phenomena such as critical blowdown flow and the level swell during blowdown from a partially water filled vessel was studied. Understanding of these phenomena is essential for the analysis of Loss-of-Coolant and steam generator steam line break accidents.

  3. Assessment study of RELAP5/MOD2, CYCLE 36. 04 based on spray start-up test for DOEL-4

    SciTech Connect

    Moeyaert, P.; Stubbe, E.

    1989-07-01

    This report presents an assessment study for the code RELAP-5 MOD-2 based on a pressurizer spray start-up test of the Doel-4 power plant. Doel-4 is a three loop WESTINGHOUSE PWR plant ordered by the EBES utility with a nominal power rating of 1000 MWe and equipped with preheater type E steam generators. A large series of commissioning tests are normally performed on new plants, of which the so called pressurizer spray and heater test (SU-PR-01) was performed on February 2nd 1985. TRACTEBEL, being the Architect-Engineer for this plant was closely involved with all start-up tests and was responsible for the final approval of the tests.

  4. SCDAP/RELAP5 Modeling of Movement of Melted Material through Porous Debris in Lower Head (Rev. 2)

    SciTech Connect

    Siefken, Larry James

    1999-10-01

    A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has been implemented into the SCDAP/RELAP5 code for the analysis of the integrity of LWR lower heads during severe accidents. The results of the model indicate that melted core plate material my permeate in about 120 s to the bottom of a 1 m deep hot porous debris bed supported by the lower head. The presence of the relocated core plate material at the bottom of the debris bed decreases the thermal resistance of the interface between the debris bed and the lower head. This report is a revision of the report with the identifier of INEEL/EXT-98-01178 REV 1, entitled "SCDAP/RELAP5 Modeling of Movement of Melted Material Through Porous Debris in Lower Head."

  5. The Impact of RELAP5 Pipe Break Flow Rates Associated With Reverse Flow Limiter Removal for Steam Generator Replacement

    SciTech Connect

    Dong Zheng; Jarvis, Julie M.; Vieira, Allen T.

    2006-07-01

    Pipe break flow rates are calculated for a main feedwater line break (FWLB) in the main steam valve vault (MSVV) for a PWR Steam Generator Replacement (SGR). A reverse flow limiter is installed in the original steam generator (OSG) feedwater nozzle to limit the blowdown flowrate in the event of a postulated FWLB. This feature is not incorporated in the replacement steam generator (RSG) design. The change in RSG nozzle design in conjunction with new operating conditions results in increased FWLB mass and energy releases which can impact environmental temperatures and pressures and flooding levels. In the United States, benchmarking for safety related analyses is necessary in consideration of 10CFR50.59 requirements. RELAP5/MOD3 is used to model the pipe break flowrates for a FWLB at different break locations. The benchmark FWLB blowdown releases are larger than the OSG design basis blowdown releases due to differences in RELAP5/MOD3 versions which are found to have different algorithms for subcooled choked flow. The SGR FWLB blowdown release rates are determined to have minimal impact on the compartment temperature and pressure response. However, the flooding levels and associated equipment qualification are potentially impacted. Modeling techniques used to minimize the impact of the SGR blowdown releases on MSVV flooding levels include modeling flashing effects, more realistic RSG temperature distribution, inventory depletion and Auxiliary Feedwater (AFW) flow initiation time, and considering loss of offsite power scenarios. A detailed flooding hazard evaluation is needed, which considers the actual main feedwater isolation times to ensure that environmentally qualified safety related components, required to mitigate the effects of a FWLB inside the MSVV, can perform their safety function prior to being submerged. (authors)

  6. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  7. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  8. Results from a scaled reactor cavity cooling system with water at steady state

    SciTech Connect

    Lisowski, D. D.; Albiston, S. M.; Tokuhiro, A.; Anderson, M. H.; Corradini, M. L.

    2012-07-01

    We present a summary of steady-state experiments performed with a scaled, water-cooled Reactor Cavity Cooling System (RCCS) at the Univ. of Wisconsin - Madison. The RCCS concept is used for passive decay heat removal in the Next Generation Nuclear Plant (NGNP) design and was based on open literature of the GA-MHTGR, HTR-10 and AVR reactor. The RCCS is a 1/4 scale model of the full scale prototype system, with a 7.6 m structure housing, a 5 m tall test section, and 1,200 liter water storage tank. Radiant heaters impose a heat flux onto a three riser tube test section, representing a 5 deg. radial sector of the actual 360 deg. RCCS design. The maximum heat flux and power levels are 25 kW/m{sup 2} and 42.5 kW, and can be configured for variable, axial, or radial power profiles to simulate prototypic conditions. Experimental results yielded measurements of local surface temperatures, internal water temperatures, volumetric flow rates, and pressure drop along the test section and into the water storage tank. The majority of the tests achieved a steady state condition while remaining single-phase. A selected number of experiments were allowed to reach saturation and subsequently two-phase flow. RELAP5 simulations with the experimental data have been refined during test facility development and separate effects validation of the experimental facility. This test series represents the completion of our steady-state testing, with future experiments investigating normal and off-normal accident scenarios with two-phase flow effects. The ultimate goal of the project is to combine experimental data from UW - Madison, UI, ANL, and Texas A and M, with system model simulations to ascertain the feasibility of the RCCS as a successful long-term heat removal system during accident scenarios for the NGNP. (authors)

  9. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  10. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    SciTech Connect

    Wulff, W.

    1990-01-01

    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 in 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.

  11. Ultrasonic inspection of reactor systems

    SciTech Connect

    Majzlik, E.J. Jr.

    1989-01-01

    The subject of this presentation is ultrasonic inspection of reactor systems. This paper describes two current programs underway at Savannah River Site which provide state-of-the-art ultrasonic inspections of weld heat-affected zones in the primary cooling loop of the Savannah River Site reactors. It also describes the automated remote inspection equipment being developed and employed; briefly describe the procedures being used; and give you a general idea of the future direction of two major programs: Moderator Piping Inspection Program and the Reactor Tank Wall Weld Inspection Program. The objective of these programs is to provide inspection techniques to more fully determine the condition of the reactor primary system and provide data for prediction of maintenance needs and remaining service life. Detection and sizing of intergranular stress corrosion cracking is the focus of these programs.

  12. Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

    SciTech Connect

    Gonzalez Gonzalez, R.; Petruzzi, A.; D'Auria, F.; Mazzantini, O.

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D{sup C} system code. Within the framework of the third Agreement 'NA-SA - Univ. of Pisa' a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions. (authors)

  13. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  14. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  15. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  16. Analysis of N Reactor anticipated transients without scram

    SciTech Connect

    Sathyanarayana, K. ); Ogden, D.M. )

    1989-07-01

    This report documents the N Reactor thermal-hydraulic system response for the anticipated transients without SCRAM (ATWS). The N Reactor is a graphite-moderated, water-cooled pressure-tube reactor. The primary coolant is pumped through 1003 horizontal pressure-tubes which contain two concentric tubular metallic fuel elements. The calculations were performed using RELAP5/MOD2 computer code. RELAP5/MOD2 is a best-estimate, two-phase, full six-equation, nonequilibrium, and nonhomogeneous thermal-hydraulic computer code designed for pressurized water reactor transients. A point-reactor kinetics model with reactivity feedback was employed to compute the power generated in the reactor core. The transients simulated were 200% double-ended guillotine break in the cold-leg manifold and the station blackout with the simultaneous failure of SCRAM and emergency core cooling system (ECCS) and the loss of ac power. In all the transients, the loss of high pressure injection system (HPI), the complete loss of graphite and shield cooling system (GSCS) and the isolation of steam generator secondary were included. Also, the ATWS transient, due to the 200% guillotine break of the cold-leg manifold assuming the availability of ECCS from three high-lift diesel pumps operating at full capacity and the loss of ac power, was analyzed. These three transients were first evaluated using the void feedback reactivity having an absolute level of {minus}167 mk corresponding to an infinite lattice MK IV fuel. Then these transients were reevaluated using a more realistic void feedback reactivity having an absolute level of {minus}70 mk. The report includes brief descriptions of the N Reactor, the computer code, the thermal-hydraulic plant model used in the simulations and the discussion of results. 12 refs., 113 figs., 13 tabs.

  17. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  18. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E.

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  19. Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981. [PWR

    SciTech Connect

    Saha, P; Jo, J H; Neymotin, L; Rohatgi, U S; Slovik, G

    1982-12-01

    This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented.

  20. Implementation of a New DTSTEP Algorithm for use in RELAP5-3D and PVMEXEC Completion Report

    SciTech Connect

    Dr. George L Mesina

    2010-12-01

    The PVM Coupling methodology for decomposing a complex model into domains onto which individual programs may be applied has proven effective for solving many multi-physics problems. There have been, from the outset, some detailed and/or long-running models that cause the process to fail. This project addressed the PVM coupling issues surrounding the DTSTEP subroutines on RELAP5-3D and PVMEXEC. Some 25 errors are listed in Tables 1 and 18 and in Section 11. These arise from deficiencies in the floating point calculation and testing of time steps, cumulative time, and time targets. The algorithmic replacement of floating point control of these items with integer based timestepping was developed and implemented. The result of the first phase, undertaken by the INL was that all but three of these issues were resolved. Moreover, two conceptual errors in DTSTEP that were not PVM coupling related were discovered and solved. The final, and most difficult three PVM Bettis User Problems, were solved during the Bettis phase of development and debugging. In 8 months since the conclusion of the project, no further DTSTEP related PVM Coupling errors have been reported.

  1. Plasma reactor waste management systems

    NASA Technical Reports Server (NTRS)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  2. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  3. Comparison of RETRAN and RELAP5 models to Oyster Creek loss of feedwater transient

    SciTech Connect

    Alammar, M.A.

    1985-07-01

    The Oyster Creek Generating Station is a 1930MW(thermal) boiling water reactor 2 plant. During the past year, a program to qualify the Oyster Creek RETRAN model against plant data was in effect at GPU Nuclear. As part of this program, a major transient that occurred on May 2, 1979, was chosen for analysis comparison. While operating at 100% power, a spurious high-pressure scram occurred, coupled with a simultaneous trip of the recirculation pumps. Other events resulted in a loss of feedwater flow and the inadvertent closure, by the operator, of the recirculation pump discharge valves, which limited recirculation flow to only five 0.0508-m (2-in.) bypass lines. The operator proceeded to isolate the vessel and use the emergency condensers for decay heat removal until feed flow was restored 45 min later. The plant RETRAN model was benchmarked against this transient for the first 45 min, using 39 volumes, 54 junctions, 25 heat conductors, and a bubble rise model for the separator/upper downcomer regions. The RETRAN results showed good agreement with plant data for downcomer level and dome pressure. The unique coupling between the downcomer and core zone liquid levels during the cyclic operation of the emergency condensers was simulated quite well. The use of the bubble rise model for the separator/ upper downcomer, however, resulted in a higher dome pressure given by RETRAN, which is believed to be due to the 100% separation efficiency of the model as compared to the degraded separator efficiencies at offoptimum operating conditions. The fuel zone liquid level was an outstanding issue at the time where a conservative simple calculation showed that the core remained covered during the transient. The RETRAN model confirmed that, but also showed that the fuel zone liquid mass during the transient was more than that at steady state.

  4. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    system scaling analysis, design parameters were obtained and designs of the compact modular 200 MWe SBWR and the full size 1200 MWe SBWR were developed. These reactors are provided with passive safety systems. A new passive vacuum breaker check valve was designed to replace the mechanical vacuum beaker check valve. The new vacuum breaker check valve was based on a hydrostatic head, and was fail safe. The performance of this new valve was evaluated both by the thermal-hydraulic code RELAP5 and by the experiments in a scaled SBWR facility, PUMA. In the core neutronic design a core depletion model was implemented to PARCS code. A lattice design for the SBWR fuel assemblies was performed. Design improvements were made to the neutronics/thermal-hydraulics models of SBWR-200 and SBWR-1200, and design analyses of these reactors were performed. The design base accident analysis and evaluation of all the passive safety systems were completed as scheduled in tasks 4 and 5. Initial conditions for the small break loss of coolant accidents (LOCA) and large break LOCA using REALP5 code were obtained. Small and large break LOCA tests were performed and the data was analyzed. An anticipated transient with scram was simulated using the RELAP5 code for SBWR-200. The transient considered was an accidental closure of the main steam isolation valve (MSIV), which was considered to be the most significant transient. The evaluation of the RELAP5 code against experimental data for SBWR-1200 was completed. In task 6, the instability analysis for the three SBWR designs (SBWR-1200, SBWR-600 and SBWR-200) were simulated for start-up transients and the results were similar. Neither the geysering instability, nor the loop type instability was predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instabilities by

  5. A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

    SciTech Connect

    Analytis, G.Th.

    1995-09-01

    As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

  6. RELAP5-3D Developmental Assessment: Comparison of Versions 4.3.4i and 4.2.1i

    SciTech Connect

    Bayless, Paul David

    2015-10-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.3.4i and 4.2.1i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  7. RELAP5-3D Developmental Assessment: Comparison of Versions 4.0.3is and 2.4.2is

    SciTech Connect

    Paul D. Bayless

    2012-09-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.0.3is and 2.4.2is. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  8. RELAP5-3D Developmental Assessment: Comparison of Versions 4.2.1i and 4.1.3i

    SciTech Connect

    Paul D. Bayless

    2014-06-01

    Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.2.1i and 4.1.3i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

  9. Assessment of RELAP5/MOD2, Cycle 36-04 using LOFT (Loss of Fluid Test) Large Break Experiment L2-5

    SciTech Connect

    Bank, Young Seok; Lee, Sang Yong; Kim, Hho-Jung . Korea Nuclear Safety Center)

    1990-04-01

    The LOFT L2-5 LBLOCA Experiment was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability to predict the phenomena in LBLOCA. One base case calculation and three cases of different nodalizations were carried out. The effect of different nodalization was studied in the area of the downcomer and core. For a sensitivity study, another calculation was executed using an updated version of RELAP5/MOD2 Cycle 36.04. A Split downcomer with one crossflow junction and two core channels were found to be effective in describing the ECC bypass and hot channel behavior. And the updated version was found to be effective in overcoming the code deficiency in the interfacial friction and reflood quenching. 11 refs., 55 figs., 10 tabs.

  10. Thermal-hydraulic modeling needs for passive reactors

    SciTech Connect

    Kelly, J.M.

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  11. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  12. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  13. Rapid starting methanol reactor system

    DOEpatents

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  14. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  15. Decay heat removal in GEN IV gas cooled fast reactors.

    SciTech Connect

    Cheng, L. Y.; Wei, T. Y. C.

    2009-08-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  16. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  17. Reactor vessel annealing system

    DOEpatents

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  18. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  19. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  20. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  1. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  2. Thermal Reactor Code System for Reactor Design and Analysis.

    SciTech Connect

    SUZUKI, TADAKAZU

    2003-04-21

    Version: 00 SRAC95 is a general purpose neutronics code system applicable to core analyses of various types of reactors, including cell calculation with burn up, core calculation for any type of thermal reactor; where core burn up calculation and fuel management were done by an auxiliary code. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications were made for nuclear data libraries and programs. In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors. SRAC95 can be used for burnup credit analysis within the ORIGEN2 and SWAT (CCC-714) code system.

  3. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOEpatents

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  4. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  5. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  6. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    SciTech Connect

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

  7. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  8. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  9. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  10. Thermal Reactor Code System for Reactor Design and Analysis.

    Energy Science and Technology Software Center (ESTSC)

    2003-04-21

    Version: 00 SRAC95 is a general purpose neutronics code system applicable to core analyses of various types of reactors, including cell calculation with burn up, core calculation for any type of thermal reactor; where core burn up calculation and fuel management were done by an auxiliary code. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications were made for nuclear data libraries and programs. In this version,more » many new functions and data are implemented to support nuclear design studies of advanced reactors. SRAC95 can be used for burnup credit analysis within the ORIGEN2 and SWAT (CCC-714) code system.« less

  11. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  12. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  13. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

    A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

  14. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  15. TREAT Reactor Control and Protection System

    SciTech Connect

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

  16. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  17. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    SciTech Connect

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  18. Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

    SciTech Connect

    Zhang, Hongbin; Zhao, Haihua; Davis, Cliff

    2008-09-01

    An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems.

  19. Cooling system for a nuclear reactor

    DOEpatents

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  20. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  1. Thermionic reactor electric propulsion system requirements.

    NASA Technical Reports Server (NTRS)

    Mondt, J. F.; Sawyer, C. D.; Schaupp, R. W.

    1972-01-01

    Results of mission analysis, system analysis and mission engineering studies to find a single nuclear electric propulsion (NEP) system which would be applicable for a broad range of unmanned outer planet missions. The NEP system studied uses an in-core nuclear thermionic reactor as the electric power source and mercury bombardment ion engines for propulsion. Many requirements, which are imposed on the NEP system by the mission, were determined from the studies in the process of trying to find a single NEP system for many missions. It is concluded that a single thermionic reactor NEP system could be useful for a broad range of unmanned outer planet missions. The thermionic reactor NEP system should have a power level in the range from 70 to 120 kWe, a system specific weight of approximately 30 kg/kWe, and a full power output capability of 20,000 hr.

  2. Scanning tunneling microscope assembly, reactor, and system

    DOEpatents

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  3. Fluidized-Bed Reactor System

    NASA Technical Reports Server (NTRS)

    Morrison, A. D.

    1985-01-01

    Gas pyrolysis in hot fluidized beds minimized by use of selectively filtered radiation and parabolic cavity. Reactor is parabolic cavity of two or more axes in which light emanating from one axis bounces off walls of cavity and passes through object axis to heat sample.

  4. REACTOR CONTROL ROD OPERATING SYSTEM

    DOEpatents

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  5. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  6. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  7. Reactor with natural convection backup cooling system

    SciTech Connect

    Koutz, S.L.; Cavallaro, L.; Kapich, D.D.

    1988-07-26

    A nuclear reactor is described comprising: a closed vessel containing a pool of secondary coolant; a reactor core located within the pool; an inlet conduit system for directing primary coolant to the lower end of the core; pumping means for effecting flow of the primary coolant into the inlet conduit system; an outlet conduit system located above the core for receiving coolant from the upper end of the core and carrying it to an exterior outlet pipe; and pressure reduction means for reducing fluid pressure in the inlet conduit system adjacent all of the one or more opening therein by locally increasing fluid velocity adjacent all of the one or more openings.

  8. Overview of Helium Cooled System Applications with RELAP at ENEA

    SciTech Connect

    Meloni, Paride; Casamirra, Maddalena

    2006-07-01

    Some years ago, within the framework of the study for the International Thermonuclear Experimental Reactor (ITER), ENEA assessed the RELAP5 code capability to simulate Helium cooled systems on the experimental data provided by the helium facility HEFUS3 (Brasimone, Italy). This activity allowed acquiring a certain experience on the limits and capabilities of the code that, in the following years, was used for a wide range of applications concerning transient and accident analysis of helium cooled systems. At first the paper summarizes the results of the assessment activity, then describes the accident analyses performed for the conceptual study of the Fusion Power Reactor with Helium-Cooled Pebble Bed Blanket (HCPBB) and the studies to support the safety design of the gas cooled Accelerator Driven System (ADS) concept. The paper concludes highlighting the RELAP adaptations to realize within the framework of the feasibility studies for a combined cycle concept of the Very High Temperature Reactor (VHTR). (authors)

  9. Control system for a small fission reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  10. Code System to Calculate Reactor Coolant System Leak Rate.

    SciTech Connect

    Bell, Pat

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report of the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.

  11. Code System to Calculate Reactor Coolant System Leak Rate.

    Energy Science and Technology Software Center (ESTSC)

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  12. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  13. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  14. International agreement report: Assessment study of RELAP-5 MOD-2 Cycle 36. 01 based on the DOEL-2 Steam Generator Tube Rupture incident of June 1979

    SciTech Connect

    Stubbe, E J

    1986-10-01

    This report presents a code assessment study based on a real plant transient that occurred at the DOEL 2 power plant in Belgium on June 25th 1979. DOEL 2 is a two-loop WESTINGHOUSE PWR plant of 392 MWe. A steam generator tube rupture occurred at the end of a heat-up phase which initiated a plant transient which required substantial operator involvement and presented many plant phenomena which are of interest for code assessment. While real plant transients are of special importance for code validation because of the elimination of code scaling uncertainties, they introduce however some uncertainties related to the specifications of the exact initial and boundary conditions which must be reconstructed from available on-line plant recordings and on-line computer diagnostics. Best estimate data have been reconstructed for an assessment study by means of the code RELAP5/MOD2/CYCLE 36.01. Because of inherent uncertainties in the plant data, the assessment work is focussed on phenomena whereby the comparison between plant data and computer data is based more on trends than on absolute values. Such approach is able to uncover basic code weaknesses and strengths which can contribute to a better understanding of the code potential.

  15. ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng,L.Y.; Ludewig, H.

    2007-08-05

    Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

  16. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect

    Not Available

    2012-09-01

    IBRF poster developed for the IBRF showcase. Describes the multifarious system designed for complex feedstock impregnation and processing. IBRF feedstock system has several unit operations combined into one robust system that provides for flexible and staged process configurations, such as spraying, soaking, low-severity pretreatment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation.

  17. The Secure, Transportable, Autonomous Reactor System

    SciTech Connect

    Brown, N.W.; Hassberger, J.A.; Smith, C.; Carelli, M.; Greenspan, E.; Peddicord, K.L.; Stroh, K.; Wade, D.C.; Hill, R.N.

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR) system is a development architecture for implementing a small nuclear power system, specifically aimed at meeting the growing energy needs of much of the developing world. It simultaneously provides very high standards for safety, proliferation resistance, ease and economy of installation, operation, and ultimate disposition. The STAR system accomplishes these objectives through a combination of modular design, factory manufacture, long lifetime without refueling, autonomous control, and high reliability.

  18. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  19. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

    2004-07-01

    This factsheet describes a research project whose goal is to design, fabricate, evaluate, and optimize a laboratory-scale microchannel reactor/heat exchanger system with thin-film or particulate catalysts for hydrogenation of o-nitroanisole and other nitro aromatic compounds, under moderate temperature and pressure.

  20. NON-CORROSIVE REACTOR FUEL SYSTEM

    DOEpatents

    Herrick, C.C.

    1962-08-14

    A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

  1. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  2. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  3. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    NASA Astrophysics Data System (ADS)

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

  4. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  5. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  6. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  7. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  8. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  9. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Helmick, H. H.; Schwenk, F. C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are: small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems.

  10. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  11. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C.

    1997-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

  12. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

  13. Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II Nuclear Power Plant

    SciTech Connect

    Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

    1993-12-01

    The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and, a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation. This transient actually occurred in the plant on June 19, 1989.

  14. RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

    NASA Astrophysics Data System (ADS)

    Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

    Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

  15. Savannah River Site reactor hardware design modification study

    SciTech Connect

    Fisher, J.E.

    1990-03-01

    A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs.

  16. Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

    SciTech Connect

    Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee

    2004-07-01

    SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)

  17. Development and Application of the Reactor Coolant On-Line Leakage Evaluation Model for Pressurized Water Reactors

    SciTech Connect

    Liang, Thomas K.S.; Hung, H.-J.; Chang, C.-J.

    2001-12-15

    With the consideration of mass unbalance, coolant shrinking, and compressibility, a model for reactor coolant leakage evaluation has been developed to quantify on-line the system leakage rate with conventional system measurements, regardless of where the leak occurs. This model has been derived from the system of total continuity, and it divides the reactor coolant system (RCS) into two regions, namely, the saturated and subcooled regions. The pressurizer is considered as a saturated region, and the remaining part of the RCS is regarded as a subcooled region. Taking the on-line measurements of the RCS including the RCS pressure, temperature, pressurizer water level, and charging and letdown flow rates, this model can directly evaluate on-line the RCS leakage rate. It is noted that this model is applicable only if the RCS remains subcooled. To verify the applicability of this model, data generated by RELAP5/MOD3 simulation and experimental measurements from the Institute of Nuclear Energy Research, Taiwan, Integral System Test Facility were adopted to assess this model. With further on-line verification against the Maanshan training simulator, this model was finally delivered to the Maanshan nuclear power plant (a three-looped Westinghouse pressurized water reactor) to assist the operator training and on-line evaluation of the RCS leakage rate. The smallest amount of leak flow that can be detected by the ROCK model is 3 gal/min.

  18. Containment system for supercritical water oxidation reactor

    SciTech Connect

    Chastagner, P.

    1991-12-31

    This invention is comprised of a system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  19. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  20. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  1. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  2. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2014-05-20

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  3. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2013-04-16

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  4. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  5. Integral reactor system and method for fuel cells

    SciTech Connect

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  6. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  7. Plasma generators, reactor systems and related methods

    DOEpatents

    Kong, Peter C.; Pink, Robert J.; Lee, James E.

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  8. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  9. Reactor component inventory system at FFTF

    SciTech Connect

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER.

  10. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  11. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  12. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  13. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  14. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-12-02

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  15. The liquid annular reactor system (LARS) propulsion

    SciTech Connect

    Maise, G.; Lazareth, O.W.; Horn, F.; Powell, J.R.; Ludewig, H. ); Lenard, R.X. )

    1991-01-05

    A new concept for very high specific impulse ({gt}2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures ({similar to}6000 K). Operating pressure is moderate ({similar to}10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of {similar to}100 cm (core L/D{approx}1.5).

  16. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

    Cheung, A.; Binderbauer, M.; Liu, F.; Qerushi, A.; Rostoker, N.; Wessel, F. J.

    2004-02-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, α-particles, expelled axially out of the system. α-particles flowing in one direction are decelerated and their energy recovered to ``power'' the system; particles expelled in the opposite direction provide thrust. Since the fusion products are charged particles, the system does not require the use of a massive-radiation shield. This paper describes a 100 MW CBFR-SPS design, including estimates for the propulsion-system parameters and masses. Specific emphasis is placed on the design of a closed-cycle, Brayton-heat engine, consisting of heat-exchangers, turbo-alternator, compressor, and finned radiators.

  17. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  18. Systems analysis of the CANDU 3 Reactor

    SciTech Connect

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  19. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  20. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  1. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  2. Weld monitor and failure detector for nuclear reactor system

    DOEpatents

    Sutton, Jr., Harry G.

    1987-01-01

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  3. Integrated systems analysis of the PIUS reactor

    SciTech Connect

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  4. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  5. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  6. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  7. Flow excursion time scales in the advanced neutron source reactor

    SciTech Connect

    Sulfredge, C.D.

    1995-04-01

    Flow excursion transients give rise to a key thermal limit for the proposed Advanced Neutron Source (ANS) reactor because its core involves many parallel flow channels with a common pressure drop. Since one can envision certain accident scenarios in which the thermal limits set by flow excursion correlations might be exceeded for brief intervals, a key objective is to determine how long a flow excursion would take to bring about a system failure that could lead to fuel damage. The anticipated time scale for flow excursions has been examined by subdividing the process into its component phenomena: bubble nucleation and growth, deceleration of the resulting two-phase flow, and finally overcoming thermal inertia to heat up the reactor fuel plates. Models were developed to estimate the time required for each individual stage. Accident scenarios involving sudden reduction in core flow or core exit pressure have been examined, and the models compared with RELAP5 output for the ANS geometry. For a high-performance reactor like the ANS, flow excursion time scales were predicted to be in the millisecond range, so that even very brief transients might lead to fuel damage. These results should prove useful whenever one must determine the time involved in any portion of a flow excursion transient.

  8. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  9. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  10. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

    Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  11. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  12. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  13. Auxiliary reactor for a hydrocarbon reforming system

    DOEpatents

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  14. Independent assessment of TRAC-PF1 (Version 7. 0), RELAP5/MOD1 (Cycle 14), and TRAC-BD1 (Version 12. 0) codes using separate-effects experiments

    SciTech Connect

    Saha, P; Jo, J H; Neymotin, L; Rohatgi, U S; Slovik, G C; Yuelys-Miksis, C

    1985-08-01

    This report presents the results of independent code assessment conducted at BNL. The TRAC-PF1 (Version 7.0) and RELAP5/MOD1 (Cycle 14) codes were assessed using the critical flow tests, level swell test, countercurrent flow limitation (CCFL) tests, post-CHF test, steam generator thermal performance tests, and natural circulation tests. TRAC-BD1 (Version 12.0) was applied only to the CCFL and post-CHF tests. The TRAC-PWR series of codes, i.e., TRAC-P1A, TRAC-PD2, and TRAC-PF1, have been gradually improved. However, TRAC-PF1 appears to need improvement in almost all categories of tests/phenomena attempted to BNL. Of the two codes, TRAC-PF1 and RELAP5/MOD1, the latter needs more improvement particularly in the areas of: CCFL, Level swell, CHF correlation and post-CHF heat transfer, and Numerical stability. For the CCFL and post-CHF tests, TRAC-BD1 provides the best overall results. However, the TRAC-BD1 interfacial shear package for the countercurrent annular flow regime needs further improvement for better prediction of CCFL phenomenon. 47 refs., 87 figs., 15 tabs.

  15. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  16. REACTOR - a Concept for establishing a System-of-Systems

    NASA Astrophysics Data System (ADS)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

  17. Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1

    SciTech Connect

    Dunn, F.E.

    1987-01-01

    The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.

  18. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  19. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  20. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K.; Alper, Naum I.

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  1. SP-100 Space Reactor Power System readiness

    NASA Astrophysics Data System (ADS)

    Josloff, A. T.; Matteo, D. N.; Bailey, H. S.

    The SP-100 Space Reactor Power System is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). An effective infrastructure of Industry, National Laboratories and Government agencies has made substantial progress since the 1988 System Design Review. Hardware development and testing has progressed to the point of resolving all key technical feasibility issues. The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety, performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996 - 1997 time frame.

  2. Experimental assessment of computer codes used for safety analysis of integral reactors

    SciTech Connect

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B.

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  3. Nuclear reactor loss of coolant protection system

    SciTech Connect

    Loose, R.A.

    1986-03-18

    A pressurized water reactor system is described of a nuclear power plant having a water storage tank for providing emergency coolant water and means provided external to the containment vessel, for use in the event of a primary loss of coolant situation, to circulate emergency water as a coolant by withdrawal through a wall of the containment vessel and return the same back through the wall of the containment vessel and passing the water through a heat exchange means prior to use as a coolant for the reactor core. The improvement described here consists of: an enslosure, the interior of which is sealed to the atmosphere, positioned adjacent to and exterior of a wall of the containment vessel; an inlet conduit, enclosed within a sealed outer casing, communicating between the interior of the containment vessel and the interior of the enclosure; an exhaust conduit, enclosed within a sealed outer casing, communicating between the interior of the enclosure and the interior of the containment vessel; a rupture disc on the inlet conduit within the enclosure, such that failure of the exhaust conduit within the enclosure will produce an increase of the pressure within the enclosure and above a predetermined pressure will fracture the rupture disc, and will circulate the coolant within the enclosure; and means within the interior of the enclosure for pumping coolant from the interior of the containment vessel through the inlet conduit, and back to the interior of the containment vessel through the exhaust conduit; whereby if either of the conduits should fail, coolant will be collected within the enclosure and sealed to the atmosphere.

  4. Colliding beam fusion reactor space propulsion system

    NASA Astrophysics Data System (ADS)

    Wessel, Frank J.; Binderbauer, Michl W.; Rostoker, Norman; Rahman, Hafiz Ur; O'Toole, Joseph

    2000-01-01

    We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The mean azimuthal velocities and temperatures of the fuel ion species are equal and the plasma current is unneutralized by the electrons. The resulting distribution functions are thermal in a moving frame of reference. The ion gyro-orbit radius is comparable to the dimensions of the confinement system, hence classical transport of the particles and energy is expected and the device is scaleable. We have analyzed the design over a range of 106-109 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, Isp~106 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameter×10-meters length, magnetic field ~7 Tesla, ion beam current ~10 A, and fuels of either D-He3,P-B11,P-Li6,D-Li6, etc. .

  5. Neutronics, reactor systems and fuels for transmutation

    SciTech Connect

    Warin, D.; Zaetta, A.; Varaine, F.; Grouiller, J. P.; Pillon, S.

    2006-07-01

    The purpose of transmutation is to reduce the mass and the radiotoxicity inventories of Minor Actinides and Long-Lived Fission Products of nuclear waste. In France, the law voted in 1991 for waste management has requested in particular the study of solutions and processes on the subject of transmutation. This review gives conclusions, at the end of the 15 year research period defined by the law, on the scientific and technical feasibility of transmutation based on neutronic aspects, reactor systems and possible fuels, from the CEA point of view. Important results are now available concerning the possibility of significantly reducing the quantity and the radiotoxicity of long-lived waste in association with a sustainable development of nuclear energy. As France has confirmed its long-term approach to nuclear energy, the most effective implementation of (recycling-)transmutation of Minor Actinides other than Plutonium and Uranium depends on the fast neutron GEN IV systems which are designed to recycle and manage their own actinides. The perspective to deploy a first series of such systems around 2040 supports the idea that progress is being made: the nuclear long-term waste would only be made up of Fission Products, whose radio toxicity considerably drops within a few hundred years. Future work will deal with pre industrial demonstrations of transmutation. (authors)

  6. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  7. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-20

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  8. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  9. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W.

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  10. An approach to model reactor core nodalization for deterministic safety analysis

    NASA Astrophysics Data System (ADS)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  11. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  12. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  13. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  14. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

    Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

    1973-01-01

    A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

  15. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  16. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  17. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Combinations for Metal Primary Reactor Containment System Components,'' in which there are no substantive... loading combinations for metal primary reactor containment system components. ADDRESSES: Please refer...

  18. Monitoring circuit for reactor safety systems

    DOEpatents

    Keefe, Donald J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

  19. Thermionic reactor systems for electric propulsion.

    NASA Technical Reports Server (NTRS)

    Mondt, J. F.

    1972-01-01

    This paper summarizes the preliminary design studies of unmanned electric propulsion spacecraft, with primary emphasis on the in-core thermionic reactor power subsystem. A 70-kWe power subsystem, with an external-fuel thermionic reactor, is shown integrated into a large L/D (about 20) electric propulsion spacecraft. The 70-kWe spacecraft is designed for launch to earth escape with a Titan-Centaur. Two 300-kWe reactor designs (external-fuel and flashlight designs from Atomic Energy Commission contracted studies) are integrated into 270-kWe electric propulsion spacecraft. The 270-kWe spacecraft are designed for launch to a 700-nmi earth orbit with a Titan III-C/7 booster. The 70-kWe thermionic reactor power subsystem is also conceptually shown as a space base power plant.

  20. Use of separate-effects experiments in verification of system thermal-hydraulics

    SciTech Connect

    Saha, P.

    1982-01-01

    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 USNRC is, therefore, sponsoring such efforts at several national laboratories including BNL. The overall assessment matrix includes separate-effects, integral and plant tests. However, this paper will focus on how the separate-effects tests can be utilized in verifying the thermal-hydraulic models that control the various stages of postulated accidents and/or transients in a LWR system.

  1. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  2. A fast shutdown system for SRS (Savannah River Site) reactors

    SciTech Connect

    Baumann, N.P.

    1990-01-01

    Power has been sharply reduced at Savannah River Site (SRS) reactors in large part to ensure that no bulk boiling occurs during hypothesized loss of coolant accidents. A fast shutdown system is essential to regain much of this lost power. Computations and experiments indicate that a He-3 injection system will serve this function. Instrumented tests of a full system are planned for early 1991 for one of the SRS reactors. 4 refs., 7 figs., 1 tab.

  3. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  4. Simulation of cold leg manifold break and station blackout revised sequences for reduced ECCS (emergency core cooling system) in the N Reactor

    SciTech Connect

    Bolander, M.A.; Chapman, J.C.; Fletcher, C.D.

    1988-02-01

    This report presents analyses of two loss-of-coolant-accident sequences of the N Reactor using the RELAPS/MOD2 computer code. RELAP5/MOD2 is a best estimate, two phase, nonhomogeneous, nonequilibrium, thermal-hydraulic, computer code designed for light water pressurized reactor transient. The N Reactor is a graphite- moderated, pressurized water reactor. The primary coolant is channeled through 1003 horizontal pressure tubes which contain two concentric tubular metallic fuel elements. The two accident sequences simulated were a double-ended guillotine break in the cold leg manifold and a station blackout. Both simulations cover the period beginning with the initiating event, either the break or the loss of ac power, to the stabilization of the core fuel element temperatures. (The station blackout calculation was carried out until the core was quenched.) The discussion presented in this report includes brief descriptions of the N Reactor, of the computer code and specific code modifications for horizontal reflood, and the computer code model used for the simulation. This discussion also presents the results and the analyses of the two calculations.

  5. Robotic system for remote maintenance of a pulsed nuclear reactor

    SciTech Connect

    Thunborg, S.

    1986-01-01

    Guidelines recently established for occupational radiation exposure specify that exposure should be as low as reasonably achievable. In conformance with these guidelines, SNL has developed a remote maintenance robot (RMR) system for use in the Sandia Pulse Reactor III (SPR III) facility. The RMR should reduce occupational radiation exposure by a factor of 4 and decrease reactor downtime. Other goals include developing a technology base for a more advanced pulse reactor and for the nuclear fuel cycle programs of the US Department of Energy and US Nuclear Regulatory Commission. The RMR has five major subsystems: (a) a chain-driven cart to bring the system into the reactor room; (b) a Puma 560 robot to perform dextrous operations; (c) a programmable turntable to orient the robot to any of the reactor's four sides; (d) a programmable overhead hoist for lifting components weighing up to 400 lb onto or off of the reactor; and (e) a supervisory control console for the system operator. Figure 1 is a schematic diagram of the turntable, hoist, and robot system in position around the SPR III reactor.

  6. Microprocessor tester for the treat upgrade reactor trip system

    SciTech Connect

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

  7. Emergency heat removal system for a nuclear reactor

    DOEpatents

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  8. Proposal of Space Reactor for Nuclear Electric Propulsion System

    NASA Astrophysics Data System (ADS)

    Nagata, Hidetaka; Nishiyama, Takaaki; Nakashima, Hideki

    Currently, the solar battery, the chemical cell, and the RI battery are used for the energy source in space. However, it is difficult for them to satisfy requirements for deep space explorations. Therefore, other electric power sources which can stably produce high electric energy output, regardless of distance from the sun, are necessary to execute such missions. Then, we here propose small nuclear reactors as power sources for deep space exploration, and consider a conceptual design of a small nuclear reactor for Nuclear Electric Propulsion System. It is found from nuclear analyses that the Gas-Cooled reactor could not meet the design requirement imposed on the core mass. On the other hand, a light water reactor is found to be a promising alternative to the Gas-Cooled reactor.

  9. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  10. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE PAGESBeta

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  11. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  12. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  13. Code System for Reactor Physics and Fuel Cycle Simulation.

    Energy Science and Technology Software Center (ESTSC)

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  14. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  15. THERMAL DESIGN OF THE ITER VACUUM VESSEL COOLING SYSTEM

    SciTech Connect

    Carbajo, Juan J; Yoder Jr, Graydon L; Kim, Seokho H

    2010-01-01

    RELAP5-3D models of the ITER Vacuum Vessel (VV) Primary Heat Transfer System (PHTS) have been developed. The design of the cooling system is described in detail, and RELAP5 results are presented. Two parallel pump/heat exchanger trains comprise the design one train is for full-power operation and the other is for emergency operation or operation at decay heat levels. All the components are located inside the Tokamak building (a significant change from the original configurations). The results presented include operation at full power, decay heat operation, and baking operation. The RELAP5-3D results confirm that the design can operate satisfactorily during both normal pulsed power operation and decay heat operation. All the temperatures in the coolant and in the different system components are maintained within acceptable operating limits.

  16. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  17. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  18. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  19. Fuel systems for compact fast space reactors

    SciTech Connect

    Cox, C.M.; Dutt, D.S.; Karnesky, R.A.

    1983-12-01

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO/sub 2/-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux.

  20. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  1. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  2. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect

    John Darrell Bess

    2009-06-01

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  3. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  4. Light Water Reactor-Pressure Vessel Surveillance project computer system

    SciTech Connect

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes.

  5. Simulation of the modified K reactor supplementary safety system

    SciTech Connect

    Paik, I.K.; Canas, L.R. ); Peterson, P.F. )

    1991-01-01

    The supplementary safety system (SSS) of the K reactor provides a second line of defense to shut down the reactor if the safety and control rods fail to scram. The SSS was originally designed to inject a neutron poison solution (ink) into the reactor tank via spargers. Recently, concerns arose that the ink inventory might run out before the ink front returned to the moderator during a loss-of-ac-power transient in which the coolant pumps coast down. Thus, a new system has been added to inject additional ink through the pump suctions so that ink will arrive in the core before depletion of the sparger ink. The MODFLOW code was developed to calculate the moderator flow distribution in Savannah River site (SRS) reactors, including the effects of inertia and stratification from buoyancy forces.

  6. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOEpatents

    Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  7. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  8. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system.

  9. Nuclear reactor heat transport system component low friction support system

    DOEpatents

    Wade, Elman E.

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  10. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  11. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  12. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  13. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, Juhani M.

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  14. NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

    DOEpatents

    Graham, C.B.; Spiewak, I.

    1960-05-31

    An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

  15. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  16. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-12

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  17. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    SciTech Connect

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  18. Reactor technology assessment and selection utilizing systems engineering approach

    NASA Astrophysics Data System (ADS)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  19. Small reactor power systems for manned planetary surface bases

    NASA Astrophysics Data System (ADS)

    Bloomfield, Harvey S.

    1987-12-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  20. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  1. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  2. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, Donald C.

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  3. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... operated from the control room. (b) The design of the vents and associated controls, instruments and power sources must conform to appendix A and appendix B of this part. (c) The vent system must be designed to... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with...

  4. Optimized Battery-Type Reactor Primary System Design Utilizing Lead

    SciTech Connect

    Yu, Yong H.; Son, Hyoung M.; Lee, Il S.; Suh, Kune Y.

    2006-07-01

    A number of small and medium size reactors are being developed worldwide as well as large electricity generation reactors for co-generation, district heating or desalination. The Seoul National University has started to develop 23 MWth BORIS (Battery Optimized Reactor Integral System) as a multi-purpose reactor. BORIS is an integral-type optimized fast reactor with an ultra long life core. BORIS is being designed to meet the Generation IV nuclear energy system goals of sustainability, safety, reliability and economics. Major features of BORIS include 20 consecutive years of operation without refueling; elimination of an intermediate heat transport loop and main coolant pump; open core without individual subassemblies; inherent negative reactivity feedback; and inherent load following capability. Its one mission is to provide incremental electricity generation to match the needs of developing nations and especially remote communities without major electrical grid connections. BORIS consists of a reactor module, heat exchanger, coolant module, guard vessel, reactor vessel auxiliary cooling system (RVACS), secondary system, containment and the seismic isolation. BORIS is designed to generate 10 MWe with the resulting thermal efficiency of 45 %. BORIS uses lead as the primary system coolant because of the inherent safety of the material. BORIS is coupled with a supercritical carbon dioxide Brayton cycle as the secondary system to gain a high cycle efficiency in the range of 45 %. The reference core consists of 757 fuel rods without assembly with an active core height of 0.8 m. The BORIS core consists of single enrichment zone composed of a Pu-MA (minor actinides)-U-N fuel and a ferritic-martensitic stainless steel clad. This study is intended to set up appropriate reactor vessel geometry by performing thermal hydraulic analysis on RVACS using computational fluid dynamics codes; to examine the liquid metal coolant behavior along the subchannels; to find out whether the

  5. System Study: Reactor Core Isolation Cooling 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  6. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  7. System Study: Reactor Core Isolation Cooling 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  8. PWR systems transient analysis: a reactor-safety perspective

    SciTech Connect

    Kennedy, M.F.; Abramson, P.B.; McDonald, T.A.

    1982-01-01

    In the simulation of transient events in large PWR reactor systems for reactor safety studies, the plant model is quite detailed and must include most of the plant components and control systems to adequately analyze the range of transients. The results discussed were calculated with the RELAP4/MOD6 code and reveal the need for the analysis to carefully review and understand the results to assure that they are not being adversely affected by the improper solution techniques or changes in models during the calculation.

  9. Dynamic analysis of gas-core reactor system

    NASA Technical Reports Server (NTRS)

    Turner, K. H., Jr.

    1973-01-01

    A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

  10. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  11. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  12. Catalytic reactor system for the tritium emissions reduction facility

    SciTech Connect

    Wieneke, R.E.

    1991-12-31

    Two platinum catalyst reactor subsystems have been built for the new Tritium Emissions Reduction Facility (TERF) at Mound. The two parallel subsystems each consist of three major components: a passive conservation heat exchanger, an electric preheater, and a catalytic reactor. All subsystem components and interconnecting piping are fabricated from Inconel 625 for high temperature strength and corrosion resistance. System connections are welded for longevity and reliability. Active elements are backed up by installed spares, and the reactor catalyst is replaceable. Since double containment of tritium processing systems is an important safety concept, the entire subsystem is enclosed in a stainless steel glovebox. Careful planning during the design phase created thermal isolation from the glovebox, and the ability to translate the entire subsystem from the glovebox for major maintenance. 4 refs.

  13. Catalytic reactor system for the tritium emissions reduction facility

    SciTech Connect

    Wieneke, R.E.

    1991-01-01

    Two platinum catalyst reactor subsystems have been built for the new Tritium Emissions Reduction Facility (TERF) at Mound. The two parallel subsystems each consist of three major components: a passive conservation heat exchanger, an electric preheater, and a catalytic reactor. All subsystem components and interconnecting piping are fabricated from Inconel 625 for high temperature strength and corrosion resistance. System connections are welded for longevity and reliability. Active elements are backed up by installed spares, and the reactor catalyst is replaceable. Since double containment of tritium processing systems is an important safety concept, the entire subsystem is enclosed in a stainless steel glovebox. Careful planning during the design phase created thermal isolation from the glovebox, and the ability to translate the entire subsystem from the glovebox for major maintenance. 4 refs.

  14. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  15. ANDES Measurements for Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Plompen, A. J. M.; Hambsch, F.-J.; Kopecky, S.; Nyman, M.; Rouki, C.; Salvador Castiñeira, P.; Schillebeeckx, P.; Belloni, F.; Berthoumieux, E.; Gunsing, F.; Lampoudis, C.; Calviani, M.; Guerrero, C.; Cano-Ott, D.; Gonzalez Romero, E.; Aïche, M.; Jurado, B.; Mathieu, L.; Derckx, X.; Farget, F.; Rodrigues Tajes, C.; Bacquias, A.; Dessagne, Ph.; Kerveno, M.; Borcea, C.; Negret, A.; Colonna, N.; Goncalves, I.; Penttilä, H.; Rinta-Antila, S.; Kolhinen, V. S.; Jokinen, A.

    2014-05-01

    A significant number of new measurements was undertaken by the ANDES “Measurements for advanced reactor systems” initiative. These new measurements include neutron inelastic scattering from 23Na, Mo, Zr, and 238U, neutron capture cross sections of 238U, 241Am, neutron induced fission cross sections of 240Pu, 242Pu, 241Am, 243Am and 245Cm, and measurements that explore the limits of the surrogate technique. The latter study the feasibility of inferring neutron capture cross sections for Cm isotopes, the neutron-induced fission cross section of 238Pu and fission yields and fission probabilities through full Z and A identification in inverse kinematics for isotopes of Pu, Am, Cm and Cf. Finally, four isotopes are studied which are important to improve predictions for delayed neutron precursors and decay heat by total absorption gamma-ray spectrometry (88Br, 94Rb, 95Rb, 137I). The measurements which are performed at state-of-the-art European facilities have the ambition to achieve the lowest possible uncertainty, and to come as close as is reasonably achievable to the target uncertainties established by sensitivity studies. An overview is presented of the activities and achievements, leaving detailed expositions to the various parties contributing to the conference.

  16. Improved methodology for integral analysis of advanced reactors employing passive safety

    NASA Astrophysics Data System (ADS)

    Muftuoglu, A. Kursad

    After four decades of experience with pressurized water reactors, a new generation of nuclear plants are emerging. These advanced designs employ passive safety which relies on natural forces, such as gravity and natural circulation. The new concept of passive safety also necessitates improvement in computational tools available for best-estimate analyses. The system codes originally designed for high pressure conditions in the presence of strong momentum sources such as pumps are challenged in many ways. Increased interaction of the primary system with the containment necessitates a tool for integral analysis. This study addresses some of these concerns. An improved tool for integral analysis coupling primary system with containment calculation is also presented. The code package is based on RELAP5 and CONTAIN programs, best-estimate thermal-hydraulics code for primary system analysis and containment code for containment analysis, respectively. The suitability is demonstrated with a postulated small break loss of coolant accident analysis of Westinghouse AP600 plant. The thesis explains the details of the analysis including the coupling model.

  17. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  18. Modification of the Core Cooling System of TRIGA 2000 Reactor

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-22

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  19. Safety program considerations for space nuclear reactor systems

    SciTech Connect

    Cropp, L.O.

    1984-08-01

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

  20. Modification of the Core Cooling System of TRIGA 2000 Reactor

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  1. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.46a Acceptance criteria for reactor coolant...

  2. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  3. Pressurized water nuclear reactor system with hot leg vortex mitigator

    SciTech Connect

    Lau, L.K.S.

    1990-09-18

    This patent describes an improvement in a pressurized water nuclear reactor system having a reactor pressure vessel, at least one steam generator, a hot leg conduit for charging of hot fluid from the reactor pressure vessel to the steam generator, and at least one cold leg conduit for return of cool fluid from the steam generator back to the reactor pressure vessel. The improvement comprises a residual heat removal device wherein: the hot leg has an inside diameter D{sub 1}; a first section of residual heat removal conduit is provided, having an inside diameter D{sub 2}, a first end for receipt of fluid from the hot leg, and a second end; a second section of residual heat removal conduit is provided connected to the reactor pressure vessel; a pump interconnects the second end of the first section of residual heat removal conduit with the second section of residual heat removal conduit; and a step nozzle of an inside diameter D{sub 3} and a length L interconnects the hot leg to the first end of the first section of residual heat removal conduit, with D{sub 3}/D{sub 1} {ge} 0.55, with D{sub 3}/D{sub 2}1.9 and L/D{sub 3} {ge} 1.44.

  4. Evaluation of a Reactor On-Line Uncertainty Monitoring System

    SciTech Connect

    Edwards, Robert M.; He, Weidong

    2001-06-17

    Robust control designs were developed to better match experimental conditions available in the TRIGA reactor. A first-order weighting function is specified for each operating range in the robust design to limit the maximum tracking error. Inclusion of the performance-weighting function as the on-line filter is discussed as a possible on-line performance-monitoring method. The scheme to evaluate an on-line uncertainty monitoring system for a robust reactor controller is shown. TRIGA reactor experiments were conducted to evaluate on-line performance-monitoring techniques. It is concluded that the observed robust-control performance-monitoring characteristics can be incorporated in an on-line decision-making process to choose appropriate robust control selection and enforcement.

  5. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  6. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  7. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1{percent}Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. {copyright} {ital 1999 American Institute of Physics.}

  8. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-22

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  9. Operation of staged membrane oxidation reactor systems

    DOEpatents

    Repasky, John Michael

    2012-10-16

    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  10. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  11. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  12. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  13. Assessment of RELAP5/MOD2 against a 10% load rejection transient from 75% steady state in the Vandellos II Nuclear Power Plant

    SciTech Connect

    Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

    1993-05-01

    The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); A load rejection from 100% to 50% (CSN); A load rejection from 75% to 65% (ANV); A feedwater turbopump trip (ANV). This copy is a report of the load rejection from 75% to 65% transient simulation. This transient was one of the tests carried out in Vandellos II NPP during the startup tests.

  14. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C.; Singer, Ralph M.; Humenik, Keith E.

    1993-01-01

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  15. Neutral-beam systems for magnetic-fusion reactors

    SciTech Connect

    Fink, J. H.

    1981-08-10

    Neutral beams for magnetic fusion reactors are at an early stage of development, and require considerable effort to make them into the large, reliable, and efficient systems needed for future power plants. To optimize their performance to establish specific goals for component development, systematic analysis of the beamlines is essential. Three ion source characteristics are discussed: arc-cathode life, gas efficiency, and beam divergence, and their significance in a high-energy neutral-beam system is evaluated.

  16. D-He-3 spherical torus fusion reactor system study

    NASA Astrophysics Data System (ADS)

    Macon, William A., Jr.

    1992-04-01

    This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

  17. A passive automated personnel accountability system for reactor emergency preparedness

    SciTech Connect

    Zimmerman, R.O.; DeLisle, G.V.; Hickey, E.E.

    1988-04-01

    In 1985 a project was undertaken at the N Reactor on the Hanford Site to develop an automated personnel accountability system to ensure accountability of all personnel within 30 minutes of a site evacuation. The decision to develop such a system was made after a full-scale evacuation drill showed that the manual accountability system in use at the time was inadequate to meet the 30-minute requirement. Accountability systems at commercial nuclear power plants were evaluated, but found to be unsuitable because they were not passive, that is, they required action on part of the user for the system to work. Approximately 2500 people could be required to evacuate the 100-N Area. Therefore, a card key system or badge exchange system was judged not to be feasible. A passive accountability system was desired for N Reactor to allow personnel to enter and leave the site in a more timely manner. To meet the need for an automated accountability system at N Reactor, a special Evacuation Accountability System (EVACS) was designed and developed. The EVACS system has three basic components: the transponder, a credit card-sized device worn with the security badge; portal monitors, which are electronically activated by the transponder; and a computer information system that contains the personnel data base. Each person wearing a transponder is accounted for automatically by walking through a portal. In this paper, a description of the hardware and software will be presented, together with problems encountered and lessons learned while adapting an existing technology to this particular use. The system is currently installed and requires acceptance testing before becoming operational.

  18. The radon monitoring system in Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Chu, M. C.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Leung, J. K. C.; Leung, K. Y.; Lin, Y. C.; Luk, K. B.; Pun, C. S. J.

    2016-02-01

    We developed a highly sensitive, reliable and portable automatic system (H3) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m3. This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

  19. Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

    SciTech Connect

    Conklin, J.C. )

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

  20. Summary of space nuclear reactor power systems, 1983--1992

    SciTech Connect

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  1. A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel

    SciTech Connect

    D. S. Wendt; R. L. Bewley; W. E. Windes

    2007-06-01

    Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4 mm gas inlet orifice diameter through which particles are removed from the reactor at the completion of each run. The particles may range from 200 µm to 2 mm in diameter. Maintaining optimal gas flow rates slightly above the minimum spouting velocity throughout the duration of each run is complicated by the variation of particle size and density as conversion and/or coating reactions proceed in addition to gas composition and temperature variations. In order to achieve uniform particle coating, prevent agglomeration of the particle bed, and monitor the reaction progress, a spouted bed monitoring system was developed. The monitoring system includes a high-sensitivity, low-response time differential pressure transducer paired with a signal processing, data acquisition, and process control unit which allows for real-time monitoring and control of the spouted bed reactor. The pressure transducer is mounted upstream of the spouted bed reactor gas inlet. The gas flow into the reactor induces motion of the particles in the bed and prevents the particles from draining from the reactor due to gravitational forces. Pressure fluctuations in the gas inlet stream are generated as the particles in the bed interact with the entering gas stream. The pressure fluctuations are produced by bulk movement of the bed, generation and movement of gas bubbles through the bed, and the individual motion of particles and particle subsets in the bed. The pressure fluctuations propagate upstream to the pressure transducer where they can be monitored. Pressure fluctuation, mean differential pressure, gas flow rate, reactor

  2. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  3. The secure, transportable, autonomous reactor (STAR): a small proliferation-resistant reactor system for developing countries

    SciTech Connect

    Brown, N W; Hassberger, J A; Smith, C F

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR), is an integrated concept for a small, proliferation-resistant nuclear power system capable of meeting the growing power demands of many regions of the developing world. The STAR approach builds on earlier work investigating the features required for implementation of such a system. The STAR approach includes establishing overall system requirements, conducting research into issues common to four reactor concepts (gas, liquid metal, light water and molten salt), and defining and performing the down-selection to a preferred concept that will serve as the basis for continued development leading to an eventual prototype. The paper indicates that a number of unique and distinguishing innovations are needed to both meet the energy demands of most of the world's developing regions and address growing nuclear proliferation concerns. These technical innovations form much of the basis underlying the STAR concept and include: eliminating on-site refueling and fuel access; incorporating a systems approach to nuclear energy supply and infrastructure design, with all aspects of equipment life, fuel and waste cycles included; small unit size enabling transportability; replaceable standardized modular design; resilient and robust design concepts leading to large safety margins, high reliability and reduced maintenance; simplicity in operation with reliance on autonomous control and remote monitoring; and waste minimization and waste form optimization.

  4. Completely modular Thermionic Reactor Ion Propulsion System (TRIPS)

    NASA Technical Reports Server (NTRS)

    Peelgren, M. L.; Kikin, G. M.; Sawyer, C. D.

    1972-01-01

    The nuclear reactor powered ion propulsion system described is an advanced completely modularized system which lends itself to development of prototype and/or flight type components without the need for complete system tests until late in the development program. This modularity is achieved in all of the subsystems and components of the electric propulsion system including (1) the thermionic fuel elements, (2) the heat rejection subsystem (heat pipes), (3) the power conditioning modules, and (4) the ion thrusters. Both flashlight and external fuel type in-core thermionic reactors are considered as the power source. The thermionic fuel elements would be useful over a range of reactor power levels. Electrical heated acceptance testing in their flight configuration is possible for the external fuel case. Nuclear heated testing by sampling methods could be used for acceptance testing of flashlight fuel elements. The use of heat pipes for cooling the collectors and as a means of heat transport to the radiator allows early prototype or flight configuration testing of a small module of the heat rejection subsystem as opposed to full scale liquid metal pumps and radiators in a large vacuum chamber. The power conditioner (p/c) is arranged in modules with passive cooling.

  5. Common cause analysis of the TREAT upgrade reactor protection system

    SciTech Connect

    Page, R.J.; Kamis, G.J.; Marbach, R.A.; Mueller, C.J.

    1984-09-01

    A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect the functioning of the RSS, and to subsequently modify the design of the RSS so that the effect was minimized. A number of common-cause initiators were identified which were capable of defeating the triple redundancy feature of the reactor scram system. By means of a systematic analysis of the effect these initiators could have on the system, it was possible to identify seven necessary design and procedural modifications that would greatly reduce the probability of the reactor being run while the RSS was in a faulted condition.

  6. Monitoring system for a liquid-cooled nuclear fission reactor

    DOEpatents

    DeVolpi, Alexander

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  7. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  8. Systems and methods for dismantling a nuclear reactor

    DOEpatents

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  9. Post impact behavior of mobile reactor core containment systems

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

    1972-01-01

    The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

  10. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    NASA Astrophysics Data System (ADS)

    Choi, J. H.; Jang, H. I.; Choi, W. Q.; Choi, Y.; Jang, J. S.; Jeon, E. J.; Joo, K. K.; Kim, B. R.; Kim, H. S.; Kim, J. Y.; Kim, S. B.; Kim, S. Y.; Kim, W.; Kim, Y. D.; Ko, Y. J.; Lee, J. K.; Lim, I. T.; Pac, M. Y.; Park, I. G.; Park, J. S.; Park, R. G.; Seo, H. K.; Seo, S. H.; Shin, C. D.; Siyeon, K.; Yeo, I. S.; Yu, I.

    2016-02-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  11. Designing visual displays and system models for safe reactor operations

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  12. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  13. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    SciTech Connect

    Nathan V. Hoffer; Nolan A. Anderson; Piyush Sabharwall

    2011-08-01

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  14. Safety characteristics of a suspended-pellet fission reactor system

    NASA Astrophysics Data System (ADS)

    Kingdon, David Ross

    A new fission reactor system with passive safety characteristics to eliminate the occurrence of loss-of-coolant accidents, reduce reactivity excursion effects, and which also provides for closure of the nuclear fuel cycle through on-site spent fuel management is examined. The concept uses multi-coated fuel pellets which are suspended by an upward moving coolant in vertical columns of the reactor core and electro-refining elemental separation to remove selected fission products prior to actinide recycling. The possibility of fuel melt following a loss-of-coolant is avoided as a decrease in coolant flow results in the removal of fuel from the core through the action of gravity alone. Average fluid velocities in the columns which are necessary to suspend the pellets are calculated and found to be consistent with the necessary heat extraction to yield ˜1--10 Wth per column. The total output power of such suspended pellet-type reactors is compared to the power necessary to provide the suspending fluid flow, yielding favourable ratios of ˜102--103. The reduction of reactivity excursion tendencies is envisaged through an ablative layer of material in the pellets which sublimates at temperatures above normal operating conditions. In the event of a power or temperature increase the particles fragment and thereby change their hydrodynamic drag characteristics, thus leading to fuel removal from the core by elutriation. Comparison of nuclear-to-thermal response times and elutriation rates for limiting power transients indicate that the present design assists in reactivity excursion mitigation. Closure of the nuclear fuel cycle is attained through a spent fuel management strategy which requires only on-site storage of a fraction of the fission products produced during reactor operation. Electro-refining separation of selected fission products combined with complete actinide recycling yields no isolation of plutonium or highly enriched uranium during the procedure. The out

  15. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

  16. Autonomous Control Capabilities for Space Reactor Power Systems

    SciTech Connect

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-04

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  17. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  18. Natural circulation reactor design safety analysis

    NASA Astrophysics Data System (ADS)

    Zheng, Dong

    2001-07-01

    This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors with reasonable accuracy, while failing to reproduce some safety related local phenomena. The natural circulation process is investigated in detail using PUMA separate effect natural circulation tests. The natural circulation flow rate and heat transfer rate have been modeled analytically and numerically. The work indicates that two-phase natural circulation has enough capability to remove decay power. However, the flow instability observed in two-phase natural circulation cases seriously challenges the feasibility of natural circulation reactor design. The instability is classified as a type of density wave instability induced by flashing. A detailed stability study is performed focusing on flashing induced instability under natural circulation condition. Various flashing phenomena have been studied and a mechanistic flashing model has been proposed and improved using a relaxation method. The developed relaxation flashing model can be applied to general two-phase non-equilibrium phenomena.

  19. Fission product scrubbing system for a nuclear reactor

    SciTech Connect

    Leach, D.S.

    1986-09-09

    A fission product scrubbing system is described for a nuclear reactor including a containment building defining a containment space for accommodating reactor components, comprising (a) means defining a water tank in the containment building; (b) a dividing wall extending into the water tank for separating the water tank into a first and a second compartment; (c) means defining a collection plenum normally hermetically sealed from the containment space and the environment externally of the containment building; (d) means defining a communication passage in the dividing wall underneath the water level in the first and second compartments for maintaining communication between the water stored in the first and second compartments; (e) a standpipe extending from the containment space into the second compartment; (f) a vent pipe extending from the collection plenum into the environment externally of the containment building; and (g) a rupture disc mounted in the vent pipe for normally blocking communication between the collection plenum and the environment.

  20. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S.

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.