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.

Adaptive Nodal Transport Methods for Reactor Transient Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas Downar; E. Lewis

2005-08-31

Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Anticipatory control of xenon in a pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Impink, A.J. Jr.

1987-02-10

A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals,more » control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.« less

Thermomechanical analysis of fast-burst reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Reactor transient control in support of PFR/TREAT TUCOP experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bsebsu, F.M.; Abotweirat, F.; Elwaer, S.

2008-07-15

The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulicmore » design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)« less

Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

NASA Technical Reports Server (NTRS)

Hall, Michael L.; Doster, Joseph M.

1986-01-01

Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

On some control problems of dynamic of reactor

NASA Astrophysics Data System (ADS)

Baskakov, A. V.; Volkov, N. P.

2017-12-01

The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

Analysis of steam generator tube rupture transients with single failure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanan, Nelson A.; Garner, Patrick L.

Calculations have been performed for steady state and postulated transients in the VVR-K reactor at the Institute of Nuclear Physics (INP), Kazakhstan. (The reactor designation in Cyrillic is BBP-K; transliterating characters to English gives VVR-K but translating words gives WWR-K.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The selection of the transients considered started during working meetings and email correspondence between Argonne National Laboratory (ANL) and INP staff. In the end the transient were defined by the INP staff. Calculations were performed for the fresh low-enriched uranium (LEU) coremore » and for four subsequent cores as beryllium is added to maintain critically during the first 15 cycles. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.« less

A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

NASA Astrophysics Data System (ADS)

Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

2018-05-01

The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

Design of an ammonia closed-loop storage system in a CSP power plant with a power tower cavity receiver

NASA Astrophysics Data System (ADS)

Abdiwe, Ramadan; Haider, Markus

2017-06-01

In this study the thermochemical system using ammonia as energy storage carrier is investigated and a transient mathematical model using MATLAB software was developed to predict the behavior of the ammonia closed-loop storage system including but not limited to the ammonia solar reactor and the ammonia synthesis reactor. The MATLAB model contains transient mass and energy balances as well as chemical equilibrium model for each relevant system component. For the importance of the dissociation and formation processes in the system, a Computational Fluid Dynamics (CFD) simulation on the ammonia solar and synthesis reactors has been performed. The CFD commercial package FLUENT is used for the simulation study and all the important mechanisms for packed bed reactors are taken into account, such as momentum, heat and mass transfer, and chemical reactions. The FLUENT simulation reveals the profiles inside both reactors and compared them with the profiles from the MATLAB code.

BWR Anticipated Transients Without Scram Leading to Instability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra, A.

2013-11-10

Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2010 CFR

2010-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2011 CFR

2011-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

Self-actuated shutdown system for a commercial size LMFBR. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

Tailoring the response of Autonomous Reactivity Control (ARC) systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qvist, Staffan A.; Hellesen, Carl; Gradecka, Malwina

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, weremore » performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyd D. Christensen; Annette L. Schafer

2013-11-01

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT andmore » onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”« less

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christensen, Boyd D.; Schafer, Annette L.

2014-02-01

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT andmore » onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”« less

Alternatives Analysis for the Resumption of Transient Testing Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee Nelson

2013-11-01

An alternatives analysis was performed for resumption of transient testing. The analysis considered eleven alternatives – including both US international facilities. A screening process was used to identify two viable alternatives from the original eleven. In addition, the alternatives analysis includes a no action alternative as required by the National Environmental Policy Act (NEPA). The alternatives considered in this analysis included: 1. Restart the Transient Reactor Test Facility (TREAT) 2. Modify the Annular Core Research Reactor (ACRR) which includes construction of a new hot cell and installation of a new hodoscope. 3. No Action

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Coupling of TRAC-PF1/MOD2, Version 5.4.25, with NESTLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knepper, P.L.; Hochreiter, L.E.; Ivanov, K.N.

1999-09-01

A three-dimensional (3-D) spatial kinetics capability within a thermal-hydraulics system code provides a more correct description of the core physics during reactor transients that involve significant variations in the neutron flux distribution. Coupled codes provide the ability to forecast safety margins in a best-estimate manner. The behavior of a reactor core and the feedback to the plant dynamics can be accurately simulated. For each time step, coupled codes are capable of resolving system interaction effects on neutronics feedback and are capable of describing local neutronics effects caused by the thermal hydraulics and neutronics coupling. With the improvements in computational technology,more » modeling complex reactor behaviors with coupled thermal hydraulics and spatial kinetics is feasible. Previously, reactor analysis codes were limited to either a detailed thermal-hydraulics model with simplified kinetics or multidimensional neutron kinetics with a simplified thermal-hydraulics model. The authors discuss the coupling of the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, with the NESTLE code.« less

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

The ANAMMOX reactor under transient-state conditions: process stability with fluctuations of the nitrogen concentration, inflow rate, pH and sodium chloride addition.

PubMed

Yu, Jin-Jin; Jin, Ren-Cun

2012-09-01

The process stability of an anaerobic ammonium oxidation (ANAMMOX) was investigated in an upflow anaerobic sludge blanket reactor subjected to overloads of 2.0- to 3.0-fold increases in substrate concentrations, inflow rates lasting 12 or 24h, extreme pH levels of 4 and 10 for 12h and a 12-h 30 g l(-1) NaCl addition. During the overloads, the nitrogen removal rate improved, and the shock period was an important factor affecting the reactor performance. In the high pH condition, the reactor performance significantly degenerated; while in the low pH condition, it did not happen. The NaCl addition caused the most serious deterioration in the reactor, which took 108 h to recover and was accompanied by a stoichiometric ratio divergence. There are well correlations between the total nitrogen and the electrical conductivity which is considered to be a convenient signal for controlling and monitoring the ANAMMOX process under transient-state conditions. Copyright © 2012 Elsevier Ltd. All rights reserved.

Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, Javier; Baker, Benjamin; Wang, Yaqi

This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/kmore » $. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$$_2$$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the control rod models in MAMMOTH and adding the BISON thermo-elastic models and thermal-fluids heat transfer.« less

A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bo Wook Rhee; Chang Jun Jeong; Hye Jeong Yun

2002-07-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzingmore » the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)« less

A microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

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. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » 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. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Fast Flux Test Facility thermal and pressure transient events during Cycle 11

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahrens, D. M.

1992-03-01

This report documents the thermal and pressure transients experienced by the Reactor Heat Transport System (RHTS) during Cycle 11 which included Cycles 11A, 11B-1, 11B-2 and 11C (i.e. 4 startups and 4 shutdowns). Cycle 11 consisted of a refueling period that began on March 14, 1989 and power operation which began on May 3, 1989 and ended on October 27, 1990. Transients resulted from secondary pump starts/stops while at refueling conditions. The major causes of transients at power were five unplanned reactor scrams from 100% power and problems with Loop 2 DHX Fan Controls During 11A.

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

ENGINEERING APPLICATIONS OF ANALOG COMPUTERS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bryant, L.T.; Janicke, M.J.; Just, L.C.

1961-02-01

Six examples are given of the application of analog computers in the fields of reactor engineering, heat transfer, and dynamics: deceleration of a reactor control rod by dashpot, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback (simulation of a TREAT transient), vibrating system with two degrees of freedom, temperature distribution in a radiating fin, and temperature distribution in an irfinite slab with variable thermal properties. (D.L.C.)

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

1999-09-27

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 heatmore » 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 temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

Oscillations in the reduction of permanganate by hydrogen peroxide or by ninhydrin in a batch reactor and mixed-mode oscillations in a continuous-flow stirred tank reactor

NASA Astrophysics Data System (ADS)

Tóthová, Mária; Nagy, Arpád; Treindl, Ľudovít.

1999-01-01

The periodical reduction of permanganate by hydrogen peroxide or by ninhydrin with transient oscillations in a closed system has been observed and discussed in relation to the first two permanganate oscillators described earlier. The mixed-mode oscillations of the permanganate-H 2O 2 oscillating system in a continuous-flow stirred tank reactor have been described.

Fluidic self-actuating control assembly

DOEpatents

Grantz, Alan L.

1979-01-01

A fluidic self-actuating control assembly for use in a reactor wherein no external control inputs are required to actuate (scram) the system. The assembly is constructed to scram upon sensing either a sudden depressurization of reactor inlet flow or a sudden increase in core neutron flux. A fluidic control system senses abnormal flow or neutron flux transients and actuates the system, whereupon assembly coolant flow reverses, forcing absorber balls into the reactor core region.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmarkmore » and with the results provided by PARCS code. (authors)« less

Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

2006-06-01

Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

TREAT Reactor Control and Protection System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

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).more » 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.« less

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

Space radiation studies at the White Sands Missile Range Fast Burst Reactor

NASA Technical Reports Server (NTRS)

Delapaz, A.

1972-01-01

The operation of the White Sands Missile Range Fast Burst Reactor is discussed. Space radiation studies in radiobiology, dosimetry, and transient radiation effects on electronic systems and components are described. Proposed modifications to increase the capability of the facility are discussed.

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

THETRIS: A MICRO-SCALE TEMPERATURE AND GAS RELEASE MODEL FOR TRISO FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi; A.M. Ougouag

2011-12-01

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with sub-millimeter sized kernels formed into TRISO particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations a meso-scale, pebble and compact scale, solution provides a good approximation of the fuel temperature. Micro-scale models aremore » necessary in order to obtain accurate predictions in faster transients or when parameters internal to the TRISO are needed. Since these coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD-THERMIX-KONVEK suite of coupled codes. The code includes gas release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. The analyses show the instances when the micro-scale models improve the predictions of the fuel temperature and Doppler feedback. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can cause unexpected responses during fast transients. Nevertheless, the strong Doppler feedback forces the reactor to quickly stabilize.« less

Accident analysis of heavy water cooled thorium breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

Design criteria for a self-actuated shutdown system to ensure limitation of core damage. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deane, N.A.; Atcheson, D.B.

1981-09-01

Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times.

Analysis of boron dilution in a four-loop PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, J.G.; Sha, W.T.

1995-12-31

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plantmore » is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.« less

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr

2015-12-31

The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problemmore » are presented.« less

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Gas-core reactor power transient analysis

NASA Technical Reports Server (NTRS)

Kascak, A. F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

Trench fast reactor design using the microcomputer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

Analysis of boron dilution in a four-loop PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, J.G.; Sha, W.T.

1995-03-01

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant ismore » at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.« less

Cultural Resource Investigations for the Resumption of Transient Testing of Nuclear Fuels and Material at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, Brenda R.; Williams, Julie B.

2013-11-01

The U. S. Department of Energy (DOE) has a need to test nuclear fuels under conditions that subject them to short bursts of intense, high-power radiation called ‘transient testing’ in order to gain important information necessary for licensing new nuclear fuels for use in U.S. nuclear power plants, for developing information to help improve current nuclear power plant performance and sustainability, for improving the affordability of new generation reactors, for developing recyclable nuclear fuels, and for developing fuels that inhibit any repurposing into nuclear weapons. To meet this mission need, DOE is considering alternatives for re-use and modification of existingmore » nuclear reactor facilities to support a renewed transient testing program. One alternative under consideration involves restarting the Transient Reactor Test (TREAT) reactor located at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL) site in southeastern Idaho. This report summarizes cultural resource investigations conducted by the INL Cultural Resource Management Office in 2013 to support environmental review of activities associated with restarting the TREAT reactor at the INL. These investigations were completed in order to identify and assess the significance of cultural resources within areas of potential effect associated with the proposed action and determine if the TREAT alternative would affect significant cultural resources or historic properties that are eligible for nomination to the National Register of Historic Places. No archaeological resources were identified in the direct area of potential effects for the project, but four of the buildings proposed for modifications are evaluated as historic properties, potentially eligible for nomination to the National Register of Historic Places. This includes the TREAT reactor (building #), control building (building #), guardhouse (building #), and warehouse (building #). The proposed re-use of these historic properties is consistent with original missions related to nuclear reactor testing and is expected to result in no adverse effects to their historic significance. Cultural resource investigations also involved communication with representatives from the Shoshone-Bannock Tribes to characterize cultural resources of potential tribal concern. This report provides a summary of the cultural resources inventoried and assessed within the defined areas of potential effect for the resumption of transient testing at the INL. Based on these analyses, proposed activities would have no adverse effects on historic properties within the APEs that have been defined. Other archaeological resources and cultural resources of potential concern to the Shoshone-Bannock Tribes and others that are located near the APEs are also discussed with regard to potential indirect impacts. The report concludes with general recommendations for measures to reduce impacts to all identified resources.« less

The effectiveness of using the calculated braking current for longitudinal differential protection of 110 - 750 kV shunt reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vdovin, S. A.; Shalimov, A. S.

2013-05-15

The use of the function of effective current braking of the longitudinal differential protection of shunt reactors to offset current surges, which enables the sensitivity of differential protection to be increased when there are short circuits with low damage currents, is considered. It is shown that the use of the calculated braking characteristic enables the reliability of offset protection from transients to be increased when the reactor is connected, which is accompanied by the flow of asymmetric currents containing an aperiodic component.

TREAT neutron-radiography facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, L.J.

1981-01-01

The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists.

Dynamic biogas upgrading based on the Sabatier process: thermodynamic and dynamic process simulation.

PubMed

Jürgensen, Lars; Ehimen, Ehiaze Augustine; Born, Jens; Holm-Nielsen, Jens Bo

2015-02-01

This study aimed to investigate the feasibility of substitute natural gas (SNG) generation using biogas from anaerobic digestion and hydrogen from renewable energy systems. Using thermodynamic equilibrium analysis, kinetic reactor modeling and transient simulation, an integrated approach for the operation of a biogas-based Sabatier process was put forward, which was then verified using a lab scale heterogenous methanation reactor. The process simulation using a kinetic reactor model demonstrated the feasibility of the production of SNG at gas grid standards using a single reactor setup. The Wobbe index, CO2 content and calorific value were found to be controllable by the H2/CO2 ratio fed the methanation reactor. An optimal H2/CO2 ratio of 3.45-3.7 was seen to result in a product gas with high calorific value and Wobbe index. The dynamic reactor simulation verified that the process start-up was feasible within several minutes to facilitate surplus electricity use from renewable energy systems. Copyright © 2014 Elsevier Ltd. All rights reserved.

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2012 CFR

2012-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2013 CFR

2013-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2014 CFR

2014-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pind, C.

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives formore » heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories.« less

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

Catalytic ignition model in a monolithic reactor with in-depth reaction

NASA Technical Reports Server (NTRS)

Tien, Ta-Ching; Tien, James S.

1990-01-01

Two transient models have been developed to study the catalytic ignition in a monolithic catalytic reactor. The special feature in these models is the inclusion of thermal and species structures in the porous catalytic layer. There are many time scales involved in the catalytic ignition problem, and these two models are developed with different time scales. In the full transient model, the equations are non-dimensionalized by the shortest time scale (mass diffusion across the catalytic layer). It is therefore accurate but is computationally costly. In the energy-integral model, only the slowest process (solid heat-up) is taken as nonsteady. It is approximate but computationally efficient. In the computations performed, the catalyst is platinum and the reactants are rich mixtures of hydrogen and oxygen. One-step global chemical reaction rates are used for both gas-phase homogeneous reaction and catalytic heterogeneous reaction. The computed results reveal the transient ignition processes in detail, including the structure variation with time in the reactive catalytic layer. An ignition map using reactor length and catalyst loading is constructed. The comparison of computed results between the two transient models verifies the applicability of the energy-integral model when the time is greater than the second largest time scale of the system. It also suggests that a proper combined use of the two models can catch all the transient phenomena while minimizing the computational cost.

Transient Testing of Nuclear Fuels and Materials in the United States

NASA Astrophysics Data System (ADS)

Wachs, Daniel M.

2012-12-01

The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

UTSG-2; A theoretical model describing the transient behavior of a pressurized water reactor natural circulation U-tube steam generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hold, A.

An advanced nonlinear transient model for calculating steady-state and dynamic behaviors of characteristic parameters of a Kraftwerk Union-type vertical natural-circulation U-tube steam generator and its main steam system is presented. This model has been expanded due to the increasing need for safety-related accident research studies. It now takes into consideration the possibilities of dryout and superheating along the secondary side of the steam generator. The resulting theoretical model is the basis of the digital code UTSG-2, which can be used both by itself and in combination with other pressurized water reactor transient codes, such as ALMOD-3.4, AMOD-4, and ATHLET.

Phenomena Important in Molten Salt Reactor Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

G T-Mohr Start-up Reactivity Insertion Transient Analysis Using Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fard, Mehdi Reisi; Blue, Thomas E.; Miller, Don W.

2006-07-01

As a part of a Department of Energy-Nuclear Engineering Research Initiative (NERI) Project, we at OSU are investigating SiC semiconductor detectors as neutron power monitors for Generation IV power reactors. As a part of this project, we are investigating the power monitoring requirements for a specific type of Generation IV reactor, namely the GT-MHR. To evaluate the power monitoring requirements for the GT-MHR that are most demanding for a SiC diode power monitor, we have developed a Simulink model to study the transient behavior of the GT-MHR. In this paper, we describe the application of the Simulink code to themore » analysis of a series of Start-up Reactivity Insertion Transients (SURITs). The SURIT is considered to be a limiting protectable accident in terms of establishing the dynamic range of a SiC power monitor because of the low count rate of the detector during the start-up and absence of the reactivity feedback mechanism at the beginning of transient. The SURIT is studied with the ultimate goal of identifying combinations of 1) reactor power scram setpoints and 2) cram initiation times (the time in which a scram must be initiated once the setpoint is exceeded) for which the GT-MHR core is protected in the event of a continuous withdrawal of a control rod bank from the core from low powers. The SURIT is initiated by withdrawing a rod bank when the reactor is cold (300 K) and sub-critical at the BOEC (Beginning of Equilibrium Cycle) condition. Various initial power levels have been considered corresponding to various degrees of sub-criticality and various source strengths. An envelope of response is determined to establish which initial powers correspond to the worst case SURIT. (authors)« less

Analysis of the TREAT LEU Conceptual Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Managementmore » and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO 2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.« less

Preparation macroconstants to simulate the core of VVER-1000 reactor

NASA Astrophysics Data System (ADS)

Seleznev, V. Y.

2017-01-01

Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

DOE PAGES

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; ...

2016-10-26

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Initial verification and validation of RAZORBACK - A research reactor transient analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talley, Darren G.

2015-09-01

This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actualmore » ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.« less

Combustion engine system

NASA Technical Reports Server (NTRS)

Houseman, John (Inventor); Voecks, Gerald E. (Inventor)

1986-01-01

A flow through catalytic reactor which selectively catalytically decomposes methanol into a soot free hydrogen rich product gas utilizing engine exhaust at temperatures of 200 to 650 C to provide the heat for vaporizing and decomposing the methanol is described. The reactor is combined with either a spark ignited or compression ignited internal combustion engine or a gas turbine to provide a combustion engine system. The system may be fueled entirely by the hydrogen rich gas produced in the methanol decomposition reactor or the system may be operated on mixed fuels for transient power gain and for cold start of the engine system. The reactor includes a decomposition zone formed by a plurality of elongated cylinders which contain a body of vapor permeable, methanol decomposition catalyst preferably a shift catalyst such as copper-zinc.

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Volkov, M. S.; Gusev, Yu. P., E-mail: GusevYP@mpei.ru; Monakov, Yu. V.

The insertion of current-limiting reactors into electrical equipment operating at a voltage of 110 and 220 kV produces a change in the parameters of the transient recovery voltages at the contacts of the circuit breakers for disconnecting short circuits, which could be the reason for the increase in the duration of the short circuit, damage to the electrical equipment and losses in the power system. The results of mathematical modeling of the transients, caused by tripping of the short circuit in a reactive electric power transmission line are presented, and data are given on the negative effect of a current-limitingmore » resistor on the rate of increase and peak value of the transient recovery voltages. Methods of ensuring the standard requirements imposed on the parameters of the transient recovery voltages when using current-limiting reactors in the high-voltage electrical equipment of power plants and substations are proposed and analyzed.« less

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

DOE PAGES

Reichenberger, Michael A.; Patel, Vishal K.; Roberts, Jeremy A.; ...

2017-03-03

Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paniagua, J.; Rohatgi, U.S.; Prasad, V.

1996-10-01

RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

Advances in modelling of condensation phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra,A.

2013-11-10

A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

Complex Wall Boundary Conditions for Modeling Combustion in Catalytic Channels

NASA Astrophysics Data System (ADS)

Zhu, Huayang; Jackson, Gregory

2000-11-01

Monolith catalytic reactors for exothermic oxidation are being used in automobile exhaust clean-up and ultra-low emissions combustion systems. The reactors present a unique coupling between mass, heat, and momentum transport in a channel flow configuration. The use of porous catalytic coatings along the channel wall presents a complex boundary condition when modeled with the two-dimensional channel flow. This current work presents a 2-D transient model for predicting the performance of catalytic combustion systems for methane oxidation on Pd catalysts. The model solves the 2-D compressible transport equations for momentum, species, and energy, which are solved with a porous washcoat model for the wall boundary conditions. A time-splitting algorithm is used to separate the stiff chemical reactions from the convective/diffusive equations for the channel flow. A detailed surface chemistry mechanism is incorporated for the catalytic wall model and is used to predict transient ignition and steady-state conversion of CH4-air flows in the catalytic reactor.

Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark; Gleicher, Frederick; Ortensi, Javier

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in amore » graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.« less

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr; CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex; Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr

2014-12-15

In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity ofmore » the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.« less

On the possibility of connecting a non-operating main circulation pump with three pumps in operation without preliminary coast-down of power-generating unit No. 5 in the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Vitkovskii, I. L.; Nikonov, S. P.; Ryasnyi, S. I.

2014-02-01

The subject of this paper is a transient caused by connection of a standby loop to three operating circulation pumps at the initial reactor heat rate equal to 70% of the rated value without preliminarily reducing it to 30% of the rated level as required by the safe operation regulations. Failure of the following normal operation systems is supposed: the first- and the second-type warning protection systems, all quick-acting reducing devices releasing steam into the auxiliary manifold, the electric heaters of the pressurizer, the pressurizer injection system, the primary cooling circuit fluid makeup/blow-through systems, and the blocking systems to shut down the main circulation pump after the level in the steam generator is exceeded. In addition, it is supposed that, under transient conditions, the valves of the turbine regulation system will be in the position in which they were at the moment of the initial event until generation of the signal for positive closing of the turbine stop valves. The first signal to actuate the reactor emergency protection system (EPS) is skipped. The failure of all quick-acting reducing devices releasing steam into the atmosphere is assumed. In addition to equipment failure, at the moment when the main circulation pump is connected, the operator erroneously puts in a new setting to maintain the power allowable for four pumps in operation-in the calculations it was taken equal to 104% of the rated level at most considering the accuracy of evaluating and maintaining the reactor heat rate-and the working group of the reactor protection and control system (P&CS) starts moving upward. On reaching the set power level, the automatic reactor power regulator stops operating and the P&CS elements remain in the position in which they are at the moment. Compliance with the design safety criteria for the adopted scenario of the transient is demonstrated.

Main steam-line break core shroud loading calculations for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1995-12-31

In July 1994, the U.S. Nuclear regulatory Commission sent out Generic Letter 94-03 to all boiling water reactors in the United States, informing them of intergranular stress corrosion cracking of core shrouds found in 2 reactors. The letter directed all to perform safety analysis of the BWR units. Penn State performed scoping calculations to determine the forces experienced by the core shroud during a main-stream line break transient.

Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mouradian, E.M.

1966-02-16

A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

Numerical simulation of cavitation bubble dynamics induced by ultrasound waves in a high frequency reactor.

PubMed

Servant, G; Caltagirone, J P; Gérard, A; Laborde, J L; Hita, A

2000-10-01

The use of high frequency ultrasound in chemical systems is of major interest to optimize chemical procedures. Characterization of an open air 477 kHz ultrasound reactor shows that, because of the collapse of transient cavitation bubbles and pulsation of stable cavitation bubbles, chemical reactions are enhanced. Numerical modelling is undertaken to determine the spatio-temporal evolution of cavitation bubbles. The calculus of the emergence of cavitation bubbles due to the acoustic driving (by taking into account interactions between the sound field and bubbles' distribution) gives a cartography of bubbles' emergence within the reactor. Computation of their motion induced by the pressure gradients occurring in the reactor show that they migrate to the pressure nodes. Computed bubbles levitation sites gives a cartography of the chemical activity of ultrasound. Modelling of stable cavitation bubbles' motion induced by the motion of the liquid gives some insight on degassing phenomena.

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately modelmore » the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.« less

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, Paul R.

1994-01-01

A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

Automatic reactor control system for transient operation

NASA Astrophysics Data System (ADS)

Lipinski, Walter C.; Bhattacharyya, Samit K.; Hanan, Nelson A.

Various programmatic considerations have delayed the upgrading of the TREAT reactor and the performance of the control system is not yet experimentally verified. The current schedule calls for the upgrading activities to occur last in the calendar year 1987. Detailed simulation results, coupled with earlier validation of individual components of the control strategy in TREAT, verify the performance of the algorithms. The control system operates within the safety envelope provided by a protection system designed to ensure reactor safety under conditions of spurious reactivity additions. The approach should be directly applicable to MMW systems, with appropriate accounting of temperature rate limitations of key components and of the inertia of the secondary system components.

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perko, Z.; Gilli, L.; Lathouwers, D.

2013-07-01

Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used techniquemore » proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)« less

Posttest data analysis of FIST experimental TRAC-BD1/MOD1 power transient experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting in only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena: (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

NASA Astrophysics Data System (ADS)

Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

2018-01-01

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

EBR-II high-ramp transients under computer control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forrester, R.J.; Larson, H.A.; Christensen, L.J.

1983-01-01

During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients.

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

Performance of a composite membrane bioreactor treating toluene vapors: inocula selection, reactor performance and behavior under transient conditions.

PubMed

Kumar, Amit; Dewulf, Jo; Vercruyssen, Aline; Van Langenhove, Herman

2009-04-01

In this study, a membrane biofilm reactor performance for toluene as a model pollutant is presented. A composite membrane consisting of a porous polyacrylonitrile (PAN) support layer coated with a very thin (0.3 microm) dense polydimethylsiloxane (PDMS) top layer was used. Batch experiments were performed to select an appropriate inocula (slaughterhouse wastewater treatment sludge with a specific toluene consumption rate of 118+/-23 microg g(-1) VSS L(-1)) among the three available sources of inoculums. The maximum elimination capacity gas-side reactor volume based (EC)v and membrane based (EC)(m, max) obtained were 609 g m(-3) h(-1) and 1.2 g m(-2) h(-1) respectively, which is much higher than other membrane bioreactors. Further experiments involved the study of the membrane biofilm reactor flexibility when operational parameters as temperature, loading rate etc. were modified. In all cases, the membrane biofilm reactor showed a rapid adaptation and new steady-states were obtained within hours. Overall, the results illustrate that membrane bioreactors can potentially be a good option for treatment of air pollutants such as toluene.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brun, J.; Reynard-Carette, C.; Carette, M.

2015-07-01

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs:more » single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)« less

REVIEW OF PROPOSED METHODOLOGY FOR A RISK- INFORMED RELAXATION TO ASME SECTION XI APPENDIX G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickson, Terry L; Kirk, Mark

2010-01-01

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclearmore » industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.« less

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Lee, C. H.

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To supportmore » this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less

MODFLOW 2.0: A program for predicting moderator flow patterns

NASA Astrophysics Data System (ADS)

Peterson, P. F.; Paik, I. K.

1991-07-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in the operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors.

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

RETRAN analysis of multiple steam generator blow down caused by an auxiliary feedwater steam-line break

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1987-01-01

Analysis results for multiple steam generator blow down caused by an auxiliary feedwater steam-line break performed with the RETRAN-02 MOD 003 computer code are presented to demonstrate the capabilities of the RETRAN code to predict system transient response for verifying changes in operational procedures and supporting plant equipment modifications. A typical four-loop Westinghouse pressurized water reactor was modeled using best-estimate versus worst case licensing assumptions. This paper presents analyses performed to evaluate the necessity of implementing an auxiliary feedwater steam-line isolation modification. RETRAN transient analysis can be used to determine core cooling capability response, departure from nucleate boiling ratio (DNBR)more » status, and reactor trip signal actuation times.« less

Long-term ammonia removal in a coconut fiber-packed biofilter: analysis of N fractionation and reactor performance under steady-state and transient conditions.

PubMed

Baquerizo, Guillermo; Maestre, Juan P; Machado, Vinicius C; Gamisans, Xavier; Gabriel, David

2009-05-01

A comprehensive study of long-term ammonia removal in a biofilter packed with coconut fiber is presented under both steady-state and transient conditions. Low and high ammonia loads were applied to the reactor by varying the inlet ammonia concentration from 90 to 260 ppm(v) and gas contact times ranging from 20 to 36 s. Gas samples and leachate measurements were periodically analyzed and used for characterizing biofilter performance in terms of removal efficiency (RE) and elimination capacity (EC). Also, N fractions in the leachate were quantified to both identify the experimental rates of nitritation and nitratation and to determine the N leachate distribution. Results showed stratification in the biofilter activity and, thus, most of the NH(3) removal was performed in the lower part of the reactor. An average EC of 0.5 kg N-NH(3)m(-3)d(-1) was obtained for the whole reactor with a maximum local average EC of 1.7 kg N-NH(3)m(-3)d(-1). Leachate analyses showed that a ratio of 1:1 of ammonium and nitrate ions in the leachate was obtained throughout steady-state operation at low ammonia loads with similar values for nitritation and nitratation rates. Low nitratation rates during high ammonia load periods occurred because large amounts of ammonium and nitrite accumulated in the packed bed, thus causing inhibition episodes on nitrite-oxidizing bacteria due to free ammonia accumulation. Mass balances showed that 50% of the ammonia fed to the reactor was oxidized to either nitrite or nitrate and the rest was recovered as ammonium indicating that sorption processes play a fundamental role in the treatment of ammonia by biofiltration.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, P.R.

1994-12-27

A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheatley, P.D.; Wagner, K.C.

The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.

Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFITmore » model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.« less

Combination of upflow anaerobic sludge blanket (UASB) reactor and partial nitritation/anammox moving bed biofilm reactor (MBBR) for municipal wastewater treatment.

PubMed

Malovanyy, Andriy; Yang, Jingjing; Trela, Jozef; Plaza, Elzbieta

2015-03-01

In this study the combination of an upflow anaerobic sludge blanket (UASB) reactor and a deammonification moving bed biofilm reactor (MBBR) for mainstream wastewater treatment was tested. The competition between aerobic ammonium oxidizing bacteria (AOB) and nitrite oxidizing bacteria (NOB) was studied during a 5months period of transition from reject water to mainstream wastewater followed by a 16months period of mainstream wastewater treatment. The decrease of influent ammonium concentration led to a wash-out of suspended biomass which had a major contribution to nitrite production. Influence of a dissolved oxygen concentration and a transient anoxia mechanism of NOB suppression were studied. It was shown that anoxic phase duration has no effect on NOB metabolism recovery and oxygen diffusion rather than affinities of AOB and NOB to oxygen determine the rate of nitrogen conversion in a biofilm system. Anammox activity remained on the level comparable to reject water treatment systems. Copyright © 2014 Elsevier Ltd. All rights reserved.

Landfill Leachate Toxicity Removal in Combined Treatment with Municipal Wastewater

PubMed Central

Kalka, J.

2012-01-01

Combined treatment of landfill leachate and municipal wastewater was performed in order to investigate the changes of leachate toxicity during biological treatment. Three laboratory A2O lab-scale reactors were operating under the same parameters (Q-8.5–10 L/d; HRT-1.4–1.6 d; MLSS 1.6–2.5 g/L) except for the influent characteristic and load. The influent of reactor I consisted of municipal wastewater amended with leachate from postclosure landfill; influent of reactor II consisted of leachate collected from transient landfill and municipal wastewater; reactor III served as a control and its influent consisted of municipal wastewater only. Toxicity of raw and treated wastewater was determinted by four acute toxicity tests with Daphnia magna, Thamnocephalus platyurus, Vibrio fischeri, and Raphidocelis subcapitata. Landfill leachate increased initial toxicity of wastewater. During biological treatment, significant decline of acute toxicity was observed, but still mixture of leachate and wastewater was harmful to all tested organisms. PMID:22623882

Hodoscope Cineradiography Of Nuclear Fuel Destruction Experiments

NASA Astrophysics Data System (ADS)

De Volpi, A.

1983-08-01

Nuclear reactor safety studies have applied cineradiographic techniques to achieve key information regarding the durability of fuel elements that are subjected to destructive transients in test reactors. Beginning with its development in 1963, the fast-neutron hodoscope has recorded data at the TREAT reactor in the United States of America. Consisting of a collimator instrumented with several hundred parallel channels of detectors and associated instrumentation, the hodoscope measures fuel motion that takes place within thick-walled steel test containers. Fuel movement is determined by detecting the emission of fast neutrons induced in the test capsule by bursts of the test reactor that last from 0.3 to 30 s. The system has been designed so as to achieve under certain typical conditions( horizontal) spatial resolution less than lmm, time resolution close to lms, mass resolution below 0.1 g, with adequate dynamic range and recording duration. A variety of imaging forms have been developed to display the results of processing and analyzing recorded data.*

Estimation of transient heat flux density during the heat supply of a catalytic wall steam methane reformer

NASA Astrophysics Data System (ADS)

Settar, Abdelhakim; Abboudi, Saïd; Madani, Brahim; Nebbali, Rachid

2018-02-01

Due to the endothermic nature of the steam methane reforming reaction, the process is often limited by the heat transfer behavior in the reactors. Poor thermal behavior sometimes leads to slow reaction kinetics, which is characterized by the presence of cold spots in the catalytic zones. Within this framework, the present work consists on a numerical investigation, in conjunction with an experimental one, on the one-dimensional heat transfer phenomenon during the heat supply of a catalytic-wall reactor, which is designed for hydrogen production. The studied reactor is inserted in an electric furnace where the heat requirement of the endothermic reaction is supplied by electric heating system. During the heat supply, an unknown heat flux density, received by the reactive flow, is estimated using inverse methods. In the basis of the catalytic-wall reactor model, an experimental setup is engineered in situ to measure the temperature distribution. Then after, the measurements are injected in the numerical heat flux estimation procedure, which is based on the Function Specification Method (FSM). The measured and estimated temperatures are confronted and the heat flux density which crosses the reactor wall is determined.

Modeling moving systems with RELAP5-3D

DOE PAGES

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

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

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Autonomous Control of Nuclear Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Basher, H.

2003-10-20

A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, Anton; Sienicki, James J.

2016-01-01

Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for loadmore » following operation in regions where dry air cooling is a requirement« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard

2016-12-29

The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate themore » models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.« less

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J; Gehl, S M

1979-01-01

GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included inmore » both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.« less

Recrystallization and grain growth induced by ELMs-like transient heat loads in deformed tungsten samples

NASA Astrophysics Data System (ADS)

Suslova, A.; El-Atwani, O.; Sagapuram, D.; Harilal, S. S.; Hassanein, A.

2014-11-01

Tungsten has been chosen as the main candidate for plasma facing components (PFCs) due to its superior properties under extreme operating conditions in future nuclear fusion reactors such as ITER. One of the serious issues for PFCs is the high heat load during transient events such as ELMs and disruption in the reactor. Recrystallization and grain size growth in PFC materials caused by transients are undesirable changes in the material, since the isotropic microstructure developed after recrystallization exhibits a higher ductile-to-brittle transition temperature which increases with the grain size, a lower thermal shock fatigue resistance, a lower mechanical strength, and an increased surface roughening. The current work was focused on careful determination of the threshold parameters for surface recrystallization, grain growth rate, and thermal shock fatigue resistance under ELM-like transient heat events. Transient heat loads were simulated using long pulse laser beams for two different grades of ultrafine-grained tungsten. It was observed that cold rolled tungsten demonstrated better power handling capabilities and higher thermal stress fatigue resistance compared to severely deformed tungsten. Higher recrystallization threshold, slower grain growth, and lower degree of surface roughening were observed in the cold rolled tungsten.

Recrystallization and grain growth induced by ELMs-like transient heat loads in deformed tungsten samples

PubMed Central

Suslova, A.; El-Atwani, O.; Sagapuram, D.; Harilal, S. S.; Hassanein, A.

2014-01-01

Tungsten has been chosen as the main candidate for plasma facing components (PFCs) due to its superior properties under extreme operating conditions in future nuclear fusion reactors such as ITER. One of the serious issues for PFCs is the high heat load during transient events such as ELMs and disruption in the reactor. Recrystallization and grain size growth in PFC materials caused by transients are undesirable changes in the material, since the isotropic microstructure developed after recrystallization exhibits a higher ductile-to-brittle transition temperature which increases with the grain size, a lower thermal shock fatigue resistance, a lower mechanical strength, and an increased surface roughening. The current work was focused on careful determination of the threshold parameters for surface recrystallization, grain growth rate, and thermal shock fatigue resistance under ELM-like transient heat events. Transient heat loads were simulated using long pulse laser beams for two different grades of ultrafine-grained tungsten. It was observed that cold rolled tungsten demonstrated better power handling capabilities and higher thermal stress fatigue resistance compared to severely deformed tungsten. Higher recrystallization threshold, slower grain growth, and lower degree of surface roughening were observed in the cold rolled tungsten. PMID:25366885

Recrystallization and grain growth induced by ELMs-like transient heat loads in deformed tungsten samples.

PubMed

Suslova, A; El-Atwani, O; Sagapuram, D; Harilal, S S; Hassanein, A

2014-11-04

Tungsten has been chosen as the main candidate for plasma facing components (PFCs) due to its superior properties under extreme operating conditions in future nuclear fusion reactors such as ITER. One of the serious issues for PFCs is the high heat load during transient events such as ELMs and disruption in the reactor. Recrystallization and grain size growth in PFC materials caused by transients are undesirable changes in the material, since the isotropic microstructure developed after recrystallization exhibits a higher ductile-to-brittle transition temperature which increases with the grain size, a lower thermal shock fatigue resistance, a lower mechanical strength, and an increased surface roughening. The current work was focused on careful determination of the threshold parameters for surface recrystallization, grain growth rate, and thermal shock fatigue resistance under ELM-like transient heat events. Transient heat loads were simulated using long pulse laser beams for two different grades of ultrafine-grained tungsten. It was observed that cold rolled tungsten demonstrated better power handling capabilities and higher thermal stress fatigue resistance compared to severely deformed tungsten. Higher recrystallization threshold, slower grain growth, and lower degree of surface roughening were observed in the cold rolled tungsten.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

NASA Astrophysics Data System (ADS)

Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

2018-01-01

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

First overpower tests of metallic IFR [Integral Fast Reactor] fuel in TREAT [Transient Reactor Test Facility]: Data and analysis from tests M5, M6, and M7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bauer, T. H.; Robinson, W. R.; Holland, J. W.

1989-12-01

Results and analyses of margin to cladding failure and pre-failure axial expansion of metallic fuel are reported for TREAT in-pile transient overpower tests M5--M7. These are the first such tests on reference binary and ternary alloy fuel of the Integral Fast Reactor (IFR) concept with burnup ranging from 1 to 10 at. %. In all cases, test fuel was subjected to an exponential power rise on an 8 s period until either incipient or actual cladding failure was achieved. Objectives, designs and methods are described with emphasis on developments unique to metal fuel safety testing. The resulting database for claddingmore » failure threshold and prefailure fuel expansion is presented. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models of cladding failures and pre-failure axial expansions are described and compared with experimental results. Reported results include: temperature, flow, and pressure data from test instrumentation; fuel motion diagnostic data principally from the fast neutron hodoscope; and test remains described from both destructive and non-destructive post-test examination. 24 refs., 144 figs., 17 tabs.« less

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou

2016-06-01

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterizationmore » of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.« less

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

NASA Astrophysics Data System (ADS)

Seiler, J. M.; Rameau, B.

Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

Test prediction for the German PKL Test K5A using RELAP4/MOD6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Y.S.; Haigh, W.S.; Sullivan, L.H.

RELAP4/MOD6 is the most recent modification in the series of RELAP4 computer programs developed to describe the thermal-hydraulic conditions attendant to postulated transients in light water reactor systems. The major new features in RELAP4/MOD6 include best-estimate pressurized water reactor (PWR) reflood transient analytical models for core heat transfer, local entrainment, and core vapor superheat, and a new set of heat transfer correlations for PWR blowdown and reflood. These new features were used for a test prediction of the Kraftwerk Union three-loop PRIMAR KREISLAUF (PKL) Reflood Test K5A. The results of the prediction were in good agreement with the experimental thermalmore » and hydraulic system data. Comparisons include heater rod surface temperature, system pressure, mass flow rates, and core mixture level. It is concluded that RELAP4/MOD6 is capable of accurately predicting transient reflood phenomena in the 200% cold-leg break test configuration of the PKL reflood facility.« less

IMPLEMENTATION OF THE IMPROVED QUASI-STATIC METHOD IN RATTLESNAKE/MOOSE FOR TIME-DEPENDENT RADIATION TRANSPORT MODELLING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zachary M. Prince; Jean C. Ragusa; Yaqi Wang

Because of the recent interest in reactor transient modeling and the restart of the Transient Reactor (TREAT) Facility, there has been a need for more efficient, robust methods in computation frameworks. This is the impetus of implementing the Improved Quasi-Static method (IQS) in the RATTLESNAKE/MOOSE framework. IQS has implemented with CFEM diffusion by factorizing flux into time-dependent amplitude and spacial- and weakly time-dependent shape. The shape evaluation is very similar to a flux diffusion solve and is computed at large (macro) time steps. While the amplitude evaluation is a PRKE solve where the parameters are dependent on the shape andmore » is computed at small (micro) time steps. IQS has been tested with a custom one-dimensional example and the TWIGL ramp benchmark. These examples prove it to be a viable and effective method for highly transient cases. More complex cases are intended to be applied to further test the method and its implementation.« less

LMFBR system-wide transient analysis: the state of the art and US validation needs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

1982-01-01

This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

NASA Astrophysics Data System (ADS)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishii, Mamoru

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results andmore » models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.« less

The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

NASA Astrophysics Data System (ADS)

Cole, Christopher J. P.

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas that should be investigated. These include developing a detailed point nodel kinetic model coupled with a finite element heat transfer model, undertaking radiation protection shielding calculations in accordance with international and national regulations, and exploring the effects of advanced fuels.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Low-temperature overpressurization protection system setpoint analysis using RETRAN-02/MOD5 for Salem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dodson, R.J.; Feltus, M.A.

The low-temperature overpressurization protection system (LTOPS) is designed to protect the reactor pressure vessel (RPV) from brittle failure during startup and cooldown maneuvers in Westinghouse pressurized water reactors. For the Salem power plants, the power-operated relief valves (PORVs) mitigate pressure increases above a setpoint where an operational startup transient may put the RPV in the embrittlement fracture zone. The Title 10, Part 50, Code of Federal Regulations Appendix G limit, given by plant technical specifications, conservatively bounds the maximum pressure allowed during those transients where the RPV can suffer brittle fracture (usually below 350{degrees}F). The Appendix G limit is amore » pressure versus temperature curve that is more restrictive at lower RPV temperatures and allows for higher pressures as the temperature approaches the upper bounding fracture temperature.« less

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Maolong; Ryals, Matthew; Ali, Amir

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentallymore » investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.« less

Dynamic System Simulation of the KRUSTY Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Application of Radiation Chemistry to Some Selected Technological Issues Related to the Development of Nuclear Energy.

PubMed

Bobrowski, Krzysztof; Skotnicki, Konrad; Szreder, Tomasz

2016-10-01

The most important contributions of radiation chemistry to some selected technological issues related to water-cooled reactors, reprocessing of spent nuclear fuel and high-level radioactive wastes, and fuel evolution during final radioactive waste disposal are highlighted. Chemical reactions occurring at the operating temperatures and pressures of reactors and involving primary transients and stable products from water radiolysis are presented and discussed in terms of the kinetic parameters and radiation chemical yields. The knowledge of these parameters is essential since they serve as input data to the models of water radiolysis in the primary loop of light water reactors and super critical water reactors. Selected features of water radiolysis in heterogeneous systems, such as aqueous nanoparticle suspensions and slurries, ceramic oxides surfaces, nanoporous, and cement-based materials, are discussed. They are of particular concern in the primary cooling loops in nuclear reactors and long-term storage of nuclear waste in geological repositories. This also includes radiation-induced processes related to corrosion of cladding materials and copper-coated iron canisters, dissolution of spent nuclear fuel, and changes of bentonite clays properties. Radiation-induced processes affecting stability of solvents and solvent extraction ligands as well oxidation states of actinide metal ions during recycling of the spent nuclear fuel are also briefly summarized.

Investigation of natural circulation instability and transients in passively safe novel modular reactor

NASA Astrophysics Data System (ADS)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal-hydraulic and nuclear coupled startup transients are performed to investigate the flow instabilities at low pressure and low power conditions. Two different power ramps are chosen to study the effect of power density on the flow instability. The experimental startup transient tests show the existence of three different flow instability mechanisms during the low pressure startup transients, i.e., flashing instability, condensation induced instability, and density wave oscillations. Flashing instability in the chimney section of the test loop and density wave oscillation are the main flow instabilities observed when the system pressure is below 0.5 MPa. They show completely different type of oscillations, i.e., intermittent oscillation and sinusoidal oscillation, in void fraction profile during the startup transients. In order to perform nuclear-coupled startup transients with void reactivity feedback, the Point Kinetics model is utilized to calculate the transient power during the startup transients. In addition, the differences between the electric resistance heaters and typical fuel element are taken into account. The reactor power calculated shows some oscillations due to flashing instability during the transients. However, the void reactivity feedback does not have significant influence on the flow instability during the startup procedure for the NMR-50. Further investigation of very small power ramp on the startup transients is carried out for the thermal-hydraulic startup transients. It is found that very small power density can eliminate the flashing oscillation in the single phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. Furthermore, initially pressurized startup procedure is investigated to eliminate the main flow instabilities. The results show that the pressurized startup procedure can suppress the flashing instability at low pressure and low power conditions. In order to have a deep understanding of natural circulation flow instability, the quasi-steady tests are performed using the test facility installed with preheater and subcooler. The effects of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback are investigated in the quasi-steady state tests. The stability boundaries are determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. In order to predict the stability boundary theoretically, linear stability analysis in the frequency domain is performed at four sections of the loop. The flashing in the chimney is considered as an axially uniform heat source. The dimensionless characteristic equation of the pressure drop perturbation is obtained by considering the void fraction effect and outlet flow resistance in the chimney section. The flashing boundary shows some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium is recommended to improve the accuracy of flashing instability boundary.

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

2016-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologiesmore » for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.« less

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

2017-01-24

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

First-wall structural analysis of the self-cooled water blanket concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, D.A.; Steiner, D.; Embrechts, M.J.

1986-01-01

A novel blanket concept recently proposed utilizes water with small amounts of dissolved lithium compound as both coolant and breeder. The inherent simplicity of this idea should result in an attractive breeding blanket for fusion reactors. In addition, the available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate this concept. First-wall and blanket designs have been developed first for the tandem mirror reactor (TMR) due to the obvious advantages of this geometry. First-wall and blanket designs will also be developed for toroidal reactors. A simple plate designmore » with coolant tubes welded on the back (side away from plasma) was chosen as the first wall for the TMR application. Dimensions and materials were chosen to minimize temperature differences and thermal stresses. A finite element code (STRAW), originally developed for the analysis of core components subjected to high-pressure transients in the fast breeder program, was utilized to evaluate stresses in the first wall.« less

A simulator-based nuclear reactor emergency response training exercise.

PubMed

Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

Scaling mechanisms of vapour/plasma shielding from laser-produced plasmas to magnetic fusion regimes

NASA Astrophysics Data System (ADS)

Sizyuk, Tatyana; Hassanein, Ahmed

2014-02-01

The plasma shielding effect is a well-known mechanism in laser-produced plasmas (LPPs) reducing laser photon transmission to the target and, as a result, significantly reducing target heating and erosion. The shielding effect is less pronounced at low laser intensities, when low evaporation rate together with vapour/plasma expansion processes prevent establishment of a dense plasma layer above the surface. Plasma shielding also loses its effectiveness at high laser intensities when the formed hot dense plasma plume causes extensive target erosion due to radiation fluxes back to the surface. The magnitude of emitted radiation fluxes from such a plasma is similar to or slightly higher than the laser photon flux in the low shielding regime. Thus, shielding efficiency in LPPs has a peak that depends on the laser beam parameters and the target material. A similar tendency is also expected in other plasma-operating devices such as tokamaks of magnetic fusion energy (MFE) reactors during transient plasma operation and disruptions on chamber walls when deposition of the high-energy transient plasma can cause severe erosion and damage to the plasma-facing and nearby components. A detailed analysis of these abnormal events and their consequences in future power reactors is limited in current tokamak reactors. Predictions for high-power future tokamaks are possible only through comprehensive, time-consuming and rigorous modelling. We developed scaling mechanisms, based on modelling of LPP devices with their typical temporal and spatial scales, to simulate tokamak abnormal operating regimes to study wall erosion, plasma shielding and radiation under MFE reactor conditions. We found an analogy in regimes and results of carbon and tungsten erosion of the divertor surface in ITER-like reactors with erosion due to laser irradiation. Such an approach will allow utilizing validated modelling combined with well-designed and well-diagnosed LPP experimental studies for predicting consequences of plasma instabilities in complex fusion environment, which are of serious concern for successful energy production.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Capabilities Development for Transient Testing of Advanced Nuclear Fuels at TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Baker, C. C.; Bess, J. D.

2016-09-01

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. Accordingly, designs for several transient irradiation tests are being pursued to enable development of advanced nuclear fuels and materials. In addition to the reactor itself, the foundation for TREAT’s capabilities also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. The challenge of creating these vehicles, especially since many of the modern data needs were not historically addressed in TREAT experiment vehicles, has necessitated amore » sizeable engineering effort. This effort is currently underway and maturing rapidly. This paper summarizes the status, future plans, and rationale for TREAT experiment vehicle capabilities. Much of the current progress is focused around understanding and demonstrating the behavior of fuel design with enhanced accident tolerance in water-cooled reactors. Additionally, several related efforts are underway to facilitate transient testing in liquid sodium, inert gas, and steam environments. This paper discusses why such a variety of capabilities are needed, outlines plans to accomplish them, and describes the relationship between transient data needs and the irradiation hardware that will support the gathering of this information.« less

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, M.D.; Lombardo, N.J.; Heard, F.J.

1988-04-01

Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and formore » uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.« less

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

ORNL rod-bundle heat-transfer test data. Volume 2. Thermal-Hydraulic Test Facility experimental data report for test 3. 03. 6AR - transient film boiling in upflow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mullins, C. B.; Felde, D. K.; Sutton, A. G.

1982-04-01

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR. This test was conducted by members of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on May 21, 1980. Objective was to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.03.6AR was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.03.6AR available. Included in the report are uncertainties in the instrument responses,more » calculated mass flows, and calculated rod powers.« less

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through cladmore » melting at 1370/sup 0/C.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Lime, J.F.; Elson, J.S.

One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

Microscale Heat Conduction Models and Doppler Feedback

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawari, Ayman I.; Ougouag, Abderrafi

2015-01-22

The objective of this project is to establish an approach for providing the fundamental input that is needed to estimate the magnitude and time-dependence of the Doppler feedback mechanism in Very High Temperature reactors. This mechanism is the foremost contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic (TRISO) coated particles. Therefore, its correct prediction is essential to the conduct of safety analyses for these reactors. Since the effect is directly dependent on the actual temperature reached by the fuel during transients, the underlying phenomena of heat deposition, heat transfer and temperaturemore » rise must be correctly predicted. To achieve the above objective, this project will explore an approach that accounts for lattice effects as well as local temperature variations and the correct definition of temperature and related local effects.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat; Hummel, Andrew John; Hiruta, Hikaru

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy ofmore » full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.« less

SP-100 system thaw and start-up strategies

NASA Astrophysics Data System (ADS)

Kirpich, A.; Choe, H.

The authors review several strategies that have been considered during calendar year 1990 for SP-100 system start-up and thaw, and provide results of screening analyses. Screening studies have identified three concepts which are capable of thawing the SP-100 system: (1) a heat pipe approach in which reactor-heated heat pipes are routed to equipment enclosed within a thaw cavity; (2) a bleed-tube approach in which perforated tubes routed through ducting and components achieve thaw progressively as the result of successively thawing bleed-holes; and (3) a NaK traceline approach in which reactor-heated NaK is routed along ducting and components and achieves lithium thaw by both radiative and conductive coupling methods. Key issues have been identified in connection with pump start-up, reactor and power converter transient behavior, and hydraulic circuit stability. Pump start-up stability is accomplished by a linked-pump hydraulic arrangement.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

LBE water interaction in sub-critical reactors: First experimental and modelling results

NASA Astrophysics Data System (ADS)

Ciampichetti, A.; Agostini, P.; Benamati, G.; Bandini, G.; Pellini, D.; Forgione, N.; Oriolo, F.; Ambrosini, W.

2008-06-01

This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

In-situ material-motion diagnostics and fuel radiography in experimental reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeVolpi, A.

1982-01-01

Material-motion monitoring has become a routine part of in-pile transient reactor-safety experiments. Diagnostic systems, such as the fast-neutron hodoscope, were developed for the purpose of providing direct time-resolved data on pre-failure fuel motion, cladding-breach time and location, and post-failure fuel relocation. Hodoscopes for this purpose have been installed at TREAT and CABRI; other types of imaging systems that have been tested are a coded-aperture at ACRR and a pinhole at TREAT. Diagnostic systems that use penetrating radiation emitted from the test section can non-invasively monitor fuel without damage to the measuring instrument during the radiographic images of test sections installedmore » in the reator. Studies have been made of applications of hodoscopes to other experimental reactors, including PBF, FARET, STF, ETR, EBR-II, SAREF-STF, and DMT.« less

SAFEGUARDS REPORT FOR THE NORTHROP PULSE RADIATION FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feinauer, E.; Thomas, R.D.

1961-03-22

Ae description is given of the Northrop pulse Radiation Facility, (NPRF), which consists of a TRlGA Mark-F reactor and associated supporting equipment. The NPRF was designed to operate in the following modes: Mode 1-100 kw steady-state operation; Mode II--Pulsed operation up to a maximum transient giving a maximum measured fuel element temperature of 470 deg C, which corresponds to an energy release of about 18 Mw-sec (approximately 1.9% sigma K/ K). The movable reactor will be operated in three general areas in the pool: adjacent to the exposure room; adjacent to the beam ponts; or at intermediate positions. Based onmore » the analyses presented and operating experience with the prototype TRIGA Mark F and other TRlGA reactors, it is concluded that operation of the NPRF does not present any undue hazard to the health and safety of the operating personnel or the public. (auth)« less

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

A User Guide to PARET/ANL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Dionne, B.; Marin-Lafleche, A.

2015-01-01

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program tomore » determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

Phenomenology and modeling of particulate corrosion product behavior in Hanford N Reactor primary coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bechtold, D.B.

1983-12-31

The levels and composition of filterable corrosion products in the Hanford N Reactor Primary Loop are measurable by filtration. The suspended crud level has ranged from 0.0005 ppM to 6.482 ppM with a median 0.050 ppM. The composition approximates magnetite. The particle size distribution has been found in 31 cases to be uniformly a log normal distribution with a count median ranging from 1.10 to 2.31 microns with a median of 1.81 microns, and the geometric standard deviation ranging from 1.60 to 2.34 with a median of 1.84. An auto-correcting inline turbidimeter was found to respond to linearly to suspendedmore » crud levels over a range 0.05 to at least 6.5 ppM by direct comparison with filter sample weights. Cause of crud bursts in the primary loop were found to be power decreases. The crud transients associated with a reactor power drop, several reactor shutdowns, and several reactor startups could be modeled consistently with each other using a simple stirred-tank, first order exchange model of particulate between makeup, coolant, letdown, and loosely adherent crud on pipe walls. Over 3/10 of the average steady running particulate crud level could be accounted for by magnetically filterable particulate in the makeup feed. A simulation model of particulate transport has been coded in FORTRAN.« less

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Development of the ANL plant dynamics code and control strategies for the supercritical carbon dioxide Brayton cycle and code validation with data from the Sandia small-scale supercritical carbon dioxide Brayton cycle test loop.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2011-11-07

Significant progress has been made in the ongoing development of the Argonne National Laboratory (ANL) Plant Dynamics Code (PDC), the ongoing investigation and development of control strategies, and the analysis of system transient behavior for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles. Several code modifications have been introduced during FY2011 to extend the range of applicability of the PDC and to improve its calculational stability and speed. A new and innovative approach was developed to couple the Plant Dynamics Code for S-CO{sub 2} cycle calculations with SAS4A/SASSYS-1 Liquid Metal Reactor Code System calculations for the transient system level behavior onmore » the reactor side of a Sodium-Cooled Fast Reactor (SFR) or Lead-Cooled Fast Reactor (LFR). The new code system allows use of the full capabilities of both codes such that whole-plant transients can now be simulated without additional user interaction. Several other code modifications, including the introduction of compressor surge control, a new approach for determining the solution time step for efficient computational speed, an updated treatment of S-CO{sub 2} cycle flow mergers and splits, a modified enthalpy equation to improve the treatment of negative flow, and a revised solution of the reactor heat exchanger (RHX) equations coupling the S-CO{sub 2} cycle to the reactor, were introduced to the PDC in FY2011. All of these modifications have improved the code computational stability and computational speed, while not significantly affecting the results of transient calculations. The improved PDC was used to continue the investigation of S-CO{sub 2} cycle control and transient behavior. The coupled PDC-SAS4A/SASSYS-1 code capability was used to study the dynamic characteristics of a S-CO{sub 2} cycle coupled to a SFR plant. Cycle control was investigated in terms of the ability of the cycle to respond to a linear reduction in the electrical grid demand from 100% to 0% at a rate of 5%/minute. It was determined that utilization of turbine throttling control below 50% load improves the cycle efficiency significantly. Consequently, the cycle control strategy has been updated to include turbine throttle valve control. The new control strategy still relies on inventory control in the 50%-90% load range and turbine bypass for fine and fast generator output adjustments, but it now also includes turbine throttling control in the 0%-50% load range. In an attempt to investigate the feasibility of using the S-CO{sub 2} cycle for normal decay heat removal from the reactor, the cycle control study was extended beyond the investigation of normal load following. It was shown that such operation is possible with the extension of the inventory and the turbine throttling controls. However, the cycle operation in this range is calculated to be so inefficient that energy would need to be supplied from the electrical grid assuming that the generator could be capable of being operated in a motoring mode with an input electrical energy from the grid having a magnitude of about 20% of the nominal plant output electrical power level in order to maintain circulation of the CO{sub 2} in the cycle. The work on investigation of cycle operation at low power level will be continued in the future. In addition to the cycle control study, the coupled PDC-SAS4A/SASSYS-1 code system was also used to simulate thermal transients in the sodium-to-CO{sub 2} heat exchanger. Several possible conditions with the potential to introduce significant changes to the heat exchanger temperatures were identified and simulated. The conditions range from reactor scram and primary sodium pump failure or intermediate sodium pump failure on the reactor side to pipe breaks and valve malfunctions on the S-CO{sub 2} side. It was found that the maximum possible rate of the heat exchanger wall temperature change for the particular heat exchanger design assumed is limited to {+-}7 C/s for less than 10 seconds. Modeling in the Plant Dynamics Code has been compared with available data from the Sandia National Laboratories (SNL) small-scale S-CO{sub 2} Brayton cycle demonstration that is being assembled in a phased approach currently at Barber-Nichols Inc. and at SNL in the future. The available data was obtained with an earlier configuration of the S-CO{sub 2} loop involving only a single-turbo-alternator-compressor (TAC) instead of two TACs, a single low temperature recuperator (LTR) instead of both a LTR and a high temperature recuperator (HTR), and fewer than the later to be installed full set of electric heaters. Due to the absence of the full heating capability as well as the lack of a high temperature recuperator providing additional recuperation, the temperature conditions obtained with the loop are too low for the loop conditions to be prototypical of the S-CO{sub 2} cycle.« less

Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schafer, Annette L.; Brown, LLoyd C.; Carathers, David C.

2014-02-01

This document contains the analysis details and summary of analyses conducted to evaluate the environmental impacts for the Resumption of Transient Fuel and Materials Testing Program. It provides an assessment of the impacts for the two action alternatives being evaluated in the environmental assessment. These alternatives are (1) resumption of transient testing using the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) and (2) conducting transient testing using the Annular Core Research Reactor (ACRR) at Sandia National Laboratory in New Mexico (SNL/NM). Analyses are provided for radiologic emissions, other air emissions, soil contamination, and groundwater contamination that couldmore » occur (1) during normal operations, (2) as a result of accidents in one of the facilities, and (3) during transport. It does not include an assessment of the biotic, cultural resources, waste generation, or other impacts that could result from the resumption of transient testing. Analyses were conducted by technical professionals at INL and SNL/NM as noted throughout this report. The analyses are based on bounding radionuclide inventories, with the same inventories used for test materials by both alternatives and different inventories for the TREAT Reactor and ACRR. An upper value on the number of tests was assumed, with a test frequency determined by the realistic turn-around times required between experiments. The estimates provided for impacts during normal operations are based on historical emission rates and projected usage rates; therefore, they are bounding. Estimated doses for members of the public, collocated workers, and facility workers that could be incurred as a result of an accident are very conservative. They do not credit safety systems or administrative procedures (such as evacuation plans or use of personal protective equipment) that could be used to limit worker doses. Doses estimated for transportation are conservative and are based on transport of the bounding radiologic inventory that will be contained in any given test. The transportation analysis assumes all transports will contain the bounding inventory.« less

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Benjamin; Backman, Marie; Williams, Paul

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decisionmore » making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.« less

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

ANALYSIS OF BORON DILUTION TRANSIENTS IN PWRS.

DOE Office of Scientific and Technical Information (OSTI.GOV)

DIAMOND,D.J.BROMLEY,B.P.ARONSON,A.L.

2004-02-04

A study has been carried out with PARCS/RELAP5 to understand the consequences of hypothetical boron dilution events in pressurized water reactors. The scenarios of concern start with a small-break loss-of-coolant accident. If the event leads to boiling in the core and then the loss of natural circulation, a boron-free condensate can accumulate in the cold leg. The dilution event happens when natural circulation is re-established or a reactor coolant pump (RCP) is restarted in violation of operating procedures. This event is of particular concern in B&W reactors with a lowered-loop design and is a Generic Safety Issue for the U.S.more » Nuclear Regulatory Commission. The results of calculations with the reestablishment of natural circulation show that there is no unacceptable fuel damage. This is determined by calculating the maximum fuel pellet enthalpy, based on the three-dimensional model, and comparing it with the criterion for damage. The calculation is based on a model of a B&W reactor at beginning of the fuel cycle. If an RCP is restarted, unacceptable fuel damage may be possible in plants with sufficiently large volumes of boron-free condensate in the cold leg.« less

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Darius D.; Kraus, Adam R.; Bucknor, Matthew D.

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary ofmore » some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states. (C) 2016 Elsevier B.V. All rights reserved.« less

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show thatmore » fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.« less

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input thatmore » describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.« less

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.

1998-04-01

For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

TREAT Modeling and Simulation Strategy

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark David

2015-09-01

This report summarizes a four-phase process used to describe the strategy in developing modeling and simulation software for the Transient Reactor Test Facility. The four phases of this research and development task are identified as (1) full core transient calculations with feedback, (2) experiment modeling, (3) full core plus experiment simulation and (4) quality assurance. The document describes the four phases, the relationship between these research phases, and anticipated needs within each phase.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« less

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun

2015-07-01

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to dependmore » largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.« less

Transient disturbance of engineered ZnO nanoparticles enhances the resistance and resilience of anammox process in wastewater treatment.

PubMed

Zhang, Zheng-Zhe; Cheng, Ya-Fei; Xu, Lian-Zeng-Ji; Bai, Yu-Hui; Xu, Jia-Jia; Shi, Zhi-Jian; Zhang, Qian-Qian; Jin, Ren-Cun

2018-05-01

The increasing use of engineered nanoparticles (NPs) in consumer and industrial products raises concerns about their environmental impacts, but their potential influence on anaerobic ammonium oxidation (anammox) process in wastewater treatment remains unknown. In this study, the response of granule-based anammox reactor to different loads of ZnONPs was investigated. The introduction of 1-5mgL -1 ZnONPs did not affect reactor performance, but 90% of the nitrogen removal capacity was deprived by a shock of 10mgL -1 ZnONPs within 3days. Anammox activity was significantly inhibited, but no significant stimulation of intracellular reactive oxygen species (ROS) production or extracellular lactate dehydrogenase (LDH) activity was observed. The inhibition was thus mainly due to the accumulation of toxic Zn(II) ions in anammox biomass. However, the resistance and resilience of this anammox reactor to ZnONPs were enhanced by intermittent perturbations in the mode of "shock-recovery". The up-regulated abundance of Zn(II)-exporter ZntA might contribute to the enhanced resistance. In addition, these repeated transient disturbances improved the functional specificity of the anammox community despite the reduction of its diversity. Overall, these results may provide useful references for evaluating and controlling the risk of NPs to anammox process. Copyright © 2017 Elsevier B.V. All rights reserved.

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

Model predictive control of a solar-thermal reactor

NASA Astrophysics Data System (ADS)

Saade Saade, Maria Elizabeth

Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

An experimental study of ammonia borane based hydrogen storage systems

NASA Astrophysics Data System (ADS)

Deshpande, Kedaresh A.

2011-12-01

Hydrogen is a promising fuel for the future, capable of meeting the demands of energy storage and low pollutant emission. Chemical hydrides are potential candidates for chemical hydrogen storage, especially for automobile applications. Ammonia borane (AB) is a chemical hydride being investigated widely for its potential to realize the hydrogen economy. In this work, the yield of hydrogen obtained during neat AB thermolysis was quantified using two reactor systems. First, an oil bath heated glass reactor system was used with AB batches of 0.13 gram (+/- 0.001 gram). The rates of hydrogen generation were measured. Based on these experimental data, an electrically heated steel reactor system was designed and constructed to handle up to 2 grams of AB per batch. A majority of components were made of stainless-steel. The system consisted of an AB reservoir and feeder, a heated reactor, a gas processing unit and a system control and monitoring unit. An electronic data acquisition system was used to record experimental data. The performance of the steel reactor system was evaluated experimentally through batch reactions of 30 minutes each, for reaction temperatures in the range from 373 K to 430 K. The experimental data showed exothermic decomposition of AB accompanied by rapid generation of hydrogen during the initial period of the reaction. 90% of the hydrogen was generated during the initial 120 seconds after addition of AB to the reactor. At 430 K, the reaction produced 12 wt.% of hydrogen. The heat diffusion in the reactor system and the process of exothermic decomposition of AB were coupled in a two-dimensional model. Neat AB thermolysis was modeled as a global first order reactions based on Arrhenius theory. The values of equation constants were derived from curve fit of experimental data. The pre-exponential constant and the activation energy were estimated to be 4 s-1 (+/- 0.4 s-1) and 13000 J mol -1 s-1 (+/- 1050 J mol-1 s -1) respectively. The model was solved in COMSOL Multiphysics. The model was capable of simulating the transient response of the system and captured the observed trends such as the decrease in reactor temperature upon addition of AB and exothermic decomposition.

Interpretation of energy deposition data from historical operation of the transient test facility (TREAT)

DOE PAGES

DeHart, Mark D.; Baker, Benjamin A.; Ortensi, Javier

2017-07-27

The Transient Test Reactor (TREAT) at Idaho National Laboratory will resume operations in late 2017 after a 23 year hiatus while maintained in a cold standby state. Over that time period, computational power and simulation capabilities have increased substantially and now allow for new multiphysics modeling possibilities that were not practical or feasible for most of TREAT's operational history. Hence the return of TREAT to operational service provides a unique opportunity to apply state-of-the-art software and associated methods in the modeling and simulation of general three-dimensional steady state and kinetic behavior for reactor operation, and for coupling of the coremore » power transient model to experiment simulations. However, measurements taken in previous operations were intended to predict power deposition in experimental samples, with little consideration of three-dimensional core power distributions. Hence, interpretation of data for the purpose of validation of modern methods can be challenging. For the research discussed herein, efforts are described for the process of proper interpretation of data from the most recent calibration experiments performed in the core, the M8 calibration series (M8-CAL). These measurements were taken between 1990 and 1993 using a set of fission wires and test fuel pins to estimate the power deposition that would be produced in fast reactor test fuel pins during the M8 experiment series. Because of the decision to place TREAT into a standby state in 1994, the M8 series of transients were never performed. However, potentially valuable information relevant for validation is available in the M8-CAL measurement data, if properly interpreted. This article describes the current state of the process of recovery of useful data from M8-CAL measurements and quantification of biases and uncertainties to potentially apply to the validation of multiphysics methods.« less

Interpretation of energy deposition data from historical operation of the transient test facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark D.; Baker, Benjamin A.; Ortensi, Javier

The Transient Test Reactor (TREAT) at Idaho National Laboratory will resume operations in late 2017 after a 23 year hiatus while maintained in a cold standby state. Over that time period, computational power and simulation capabilities have increased substantially and now allow for new multiphysics modeling possibilities that were not practical or feasible for most of TREAT's operational history. Hence the return of TREAT to operational service provides a unique opportunity to apply state-of-the-art software and associated methods in the modeling and simulation of general three-dimensional steady state and kinetic behavior for reactor operation, and for coupling of the coremore » power transient model to experiment simulations. However, measurements taken in previous operations were intended to predict power deposition in experimental samples, with little consideration of three-dimensional core power distributions. Hence, interpretation of data for the purpose of validation of modern methods can be challenging. For the research discussed herein, efforts are described for the process of proper interpretation of data from the most recent calibration experiments performed in the core, the M8 calibration series (M8-CAL). These measurements were taken between 1990 and 1993 using a set of fission wires and test fuel pins to estimate the power deposition that would be produced in fast reactor test fuel pins during the M8 experiment series. Because of the decision to place TREAT into a standby state in 1994, the M8 series of transients were never performed. However, potentially valuable information relevant for validation is available in the M8-CAL measurement data, if properly interpreted. This article describes the current state of the process of recovery of useful data from M8-CAL measurements and quantification of biases and uncertainties to potentially apply to the validation of multiphysics methods.« less

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developedmore » at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.« less

Study on bubbly flow behavior in natural circulation reactor by thermal-hydraulic simulation tests with SF6-Gas and ethanol liquid

NASA Astrophysics Data System (ADS)

Kondo, Yoshiyuki; Suga, Keishi; Hibi, Koki; Okazaki, Toshihiko; Komeno, Toshihiro; Kunugi, Tomoaki; Serizawa, Akimi; Yoneda, Kimitoshi; Arai, Takahiro

2009-02-01

An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior in a single-channel geometry, a multi-rod-bundle one, and a horizontal-tube-bundle one on a typical natural circulation reactor system. Those experiments have clarified a) a flow regime map in a rod bundle on the transient region between bubbly and churn flow, b) three-dimensional flow behaviour in rod-bundles where inter-subassembly cross-flow occurs, c) bubble-separation behavior with consideration of reactor internal structures. The data have given analysis models for the natural circulation reactor design with good extrapolation.

Inactivation of food spoilage microorganisms by hydrodynamic cavitation to achieve pasteurization and sterilization of fluid foods.

PubMed

Milly, P J; Toledo, R T; Harrison, M A; Armstead, D

2007-11-01

Hydrodynamic cavitation is the formation of gas bubbles in a fluid due to pressure fluctuations induced by mechanical means. Various high-acid (pH < [corrected]/= 4.6) fluid foods were processed in a hydrodynamic cavitation reactor to determine if commercial sterility can be achieved at reduced processing temperatures. Sporicidal properties of the process were also tested on a low-acid (pH > [corrected] 4.6) fluid food. Fluid foods were pumped under pressure into a hydrodynamic cavitation reactor and subjected to 2 rotor speeds and flow rates to achieve 2 designated exit temperatures. Thermal inactivation kinetics were used to determine heat-induced lethality for all organisms. Calcium-fortified apple juice processed at 3000 and 3600 rpm rotor speeds on the reactor went through a transient temperature change from 20 to 65.6 or 76.7 degrees C and the total process lethality exceeded 5-log reduction of Lactobacillus plantarum and Lactobacillus sakei cells, and Zygosaccharomyces bailii cells and ascospores. Tomato juice inoculated with Bacillus coagulans spores and processed at 3000 and 3600 rpm rotor speeds endured a transient temperature from 37.8 to 93.3 or 104.4 degrees C with viable CFU reductions of 0.88 and 3.10 log cycles, respectively. Skim milk inoculated with Clostridium sporogenes putrefactive anaerobe 3679 spores and processed at 3000 or 3600 rpm rotor speeds endured a transient temperature from 48.9 to 104.4 or 115.6 degrees C with CFU reductions of 0.69 and 2.84 log cycles, respectively. Utilizing hydrodynamic cavitation to obtain minimally processed pasteurized low-acid and commercially sterilized high-acid fluid foods is possible with appropriate process considerations for different products.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

NASA Astrophysics Data System (ADS)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

NASA Astrophysics Data System (ADS)

Scarlat, Raluca Olga

This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling approach to the PB-FHR annular pebble bed core cooled by fluoride salt mixtures generated a model that is called Pod. Pod. was used to show the resilience of the PB-FHR core to generation of hot spots or cold spots, due to the effect of buoyancy on the flow and temperature distribution in the packed bed. Pod. was used to investigate the PB-FHR response to ATWS transients. Based on the functional requirements for the core, Pod. was used to generate an optimized design of the flow distribution in the core. An analysis of natural circulation loops cooled by single-phase Boussinesq fluids is presented here, in the context of reactor design that relies on natural circulation decay heat removal, and design of scaled experiments. The scaling arguments are established for a transient natural circulation loop, for loops that have long fluid residence time, and negligible contribution of fluid inertia to the momentum equation. The design of integral effects tests for the loss of forced circulation (LOFC) for PB-FHR is discussed. The special case of natural circulation decay heat removal from a pebble bed reactor was analyzed. A way to define the Reynolds number in a multi-dimensional pebble bed was identified. The scaling methodology for replicating pebble bed friction losses using an electrically resistance heated annular pipe and a needle valve was developed. The thermophysical properties of liquid fluoride salts lead to design of systems with low flow velocities, and hence long fluid residence times. A comparison among liquid coolants for the performance of steady state natural circulation heat removal from a pebble bed was performed. Transient natural circulation experimental data with simulant fluids for fluoride salts is given here. The low flow velocity and the relatively high viscosity of the fluoride salts lead to low Reynolds number flows, and a low Reynolds number in conjunction with a sufficiently high coefficient of thermal expansion makes the system susceptible to local buoyancy effects Experiments indicate that slow exchange of stagnant fluid in static legs can play a significant role in the transient response of natural circulation loops. The effect of non-linear temperature profiles on the hot or cold legs or other segments of the flow loop, which may develop during transient scenarios, should be considered when modeling the performance of natural circulation loops. The data provided here can be used for validation of the application of thermal-hydraulic systems codes to the modeling of heat removal by natural circulation with liquid fluoride salts and its simulant fluids.

Thermal and hydraulic analysis of a cylindrical blanket module design for a tokamak reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, A.Y.

1978-10-01

Various existing blanket design concepts for a tokamak fusion reactor were evaluated and assessed. These included the demonstration power reactors of ORNL, GA and others. As a result of this study, a cylindrical, modularized blanket design concept was developed. The module is a double-walled, stainless steel 316 cylinder containing liquid lithium for tritium breeding and is cooled by pressurized helium. Steady state and transient thermal conditions under normal and some off-design conditions were analyzed and presented. At the steady state reference operating point the maximum structure temperature is 452/sup 0/C at the maximum stressed location and is 495/sup 0/C atmore » the less stressed location. The coolant inlet pressure is 54.4 atm, the inlet temperature is 200/sup 0/C and the exit temperature is 435/sup 0/C. The coolant could be utilized with a helium/steam turbine power conversion system with a cycle thermal efficiency of 30.8%.« less

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

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

RELAP5 Application to Accident Analysis of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 accidentmore » 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.« less

USING CFD TO ANALYZE NUCLEAR SYSTEMS BEHAVIOR: DEFINING THE VALIDATION REQUIREMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard Schultz

2012-09-01

A recommended protocol to formulate numeric tool specifications and validation needs in concert with practices accepted by regulatory agencies for advanced reactors is described. The protocol is based on the plant type and perceived transient and accident envelopes that translates to boundary conditions for a process that gives the: (a) key phenomena and figures-of-merit which must be analyzed to ensure that the advanced plant can be licensed, (b) specification of the numeric tool capabilities necessary to perform the required analyses—including bounding calculational uncertainties, and (c) specification of the validation matrices and experiments--including the desired validation data. The result of applyingmore » the process enables a complete program to be defined, including costs, for creating and benchmarking transient and accident analysis methods for advanced reactors. By following a process that is in concert with regulatory agency licensing requirements from the start to finish, based on historical acceptance of past licensing submittals, the methods derived and validated have a high probability of regulatory agency acceptance.« less

Transient thermohydraulic heat pipe modeling

NASA Astrophysics Data System (ADS)

Hall, Michael L.; Doster, Joseph M.

Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.

1982-12-01

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency andmore » R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 bemore » 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.« less

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Swelling and gas release in oxide fuels during fast temperature transients

NASA Astrophysics Data System (ADS)

Dollins, C. C.; Jursich, M.

1982-05-01

A previously reported intergranular swelling and gas release model for oxide fuels has been modified to predict fission gas behavior during fast temperature transients. Under steady state or slowly varying conditions it has been assumed in the previous model that the pressure caused by the fission gas within the gas bubbles is in equilibrium with the surface tension of the bubbles. During a fast transient, however, net vacancy migration to the bubbles may be insufficient to maintain this equilibrium. In order to ascertain the net vacancy flow, it is necessary to model the point defect behavior in the fuel. Knowing the net flow of vacancies to the bubble and the bubble size, the bubble diffusivity can be determined and the long range migration of the gas out of the fuel can be calculated. The model has also been modified to allow release of all the gas on the grain boundaries during a fast temperature transient. The gas release predicted by the revised model shows good agreement to fast transient gas release data from an EBR-II TREAT H-3 (Transient Reactor Test Facility) test. Agreement has also been obtained between predictions using the model and gas release data obtained by Argonne National Laboratory from out-of-reactor transient heating experiments on irradiated UO 2. It was found necessary to increase the gas bubble diffusivity used in the model by a factor of thirty during the transient to provide agreement between calculations and measurements. Other workers have also found that such an increase is necessary for agreement and attribute the increased diffusivity to yielding at the bubble surface due to the increased pressure.

TREAT Transient Analysis Benchmarking for the HEU Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

2014-05-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used tomore » determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.« less

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Fuel pin cladding

DOEpatents

Vaidyanathan, Swaminathan; Adamson, Martyn G.

1986-01-01

An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1996-12-31

The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient.

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, J.E.; Knudson, D.L.; Villard, J.F.

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physicalmore » property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)« less

The RERTR Program : a status report.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 andmore » 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 generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.

1975-11-01

A mathematical model and a numerical solution scheme for thermal- hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media. (auth)

A DATA ACQUISITION SYSTEM FOR THE STUDY OF TRANSIENT RADIATION EFFECTS ON ELECTRONIC DEVICES. Paper 5 of FOURTH RADIATION EFFECTS SYMPOSIUM, SEPTEMBER 15-16, 1959, CINCINNATI, OHIO. GENERAL SESSION PAPERS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cochrane, D.O.; Graham, F.E.; Sauer, H.S.

1961-10-31

ning both the transient and permanent effects that an environment of the type created by a nuclear detonation or a pulsed reactor exerts on electronic devices, is described. The design of suitable test heads for containing the electronic devices is discussed. The design of a blockhouse for use near Ground Sero when evaluating components in a weapons environment is also discussed. (C.J.G.)

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1986-01-28

Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

Fuel pin cladding

DOEpatents

Vaidyanathan, S.; Adamson, M.G.

1983-12-16

An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

CHI Research on NSTX-U

NASA Astrophysics Data System (ADS)

Lay, W.-S.; Raman, R.; Jarboe, T. R.; Nelson, B. A.; Mueller, D.; Ebrahimi, F.; Ono, M.; Jardin, S. C.; Taylor, G.

2017-10-01

At present about 20% of the total plasma current required for sustained operation has been generated by transient CHI. The present understanding suggests that it may be possible to generate all of the needed current in a ST / tokamak using transient CHI. In such a scenario, one could transition directly from a CHI produced plasma to a non-inductively sustained plasma, without the difficult intermediate step that involves non-inductive current ramp-up. STs based on this new configuration would take advantage of evolving developments in high-temperature superconductor technology to develop a simpler design ST that relies primarily on CHI for plasma current generation. Motivated by the very good results from NSTX and HIT-II, we are examining the potential application of transient CHI for reactor configurations through these studies. (1) Study of the maximum levels of start-up currents that could be generated on NSTX-U, (2) application of a single biased electrode configuration on QUEST to protect the insulator from neutron damage in a CHI reactor installation, and (3) QUEST-like, but a double biased electrode configuration for PEGASUS and NSTX-U. Results from these on-going studies will be described. This work is supported by U.S. DOE Contracts: DE-AC02-09CH11466, DE-FG02-99ER54519 AM08, and DE-SC0006757.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

The effect of core configuration on temperature coefficient of reactivity in IRR-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bettan, M.; Silverman, I.; Shapira, M.

1997-08-01

Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is coremore » behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.« less

Neutronics Analyses of the Minimum Original HEU TREAT Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.

2014-04-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumedmore » to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.« less

Future Challenges in Heterogeneous Catalysis: Understanding Catalysts under Dynamic Reaction Conditions

PubMed Central

Kalz, Kai F.; Kraehnert, Ralph; Dvoyashkin, Muslim; Dittmeyer, Roland; Gläser, Roger; Krewer, Ulrike; Reuter, Karsten

2016-01-01

Abstract In the future, (electro‐)chemical catalysts will have to be more tolerant towards a varying supply of energy and raw materials. This is mainly due to the fluctuating nature of renewable energies. For example, power‐to‐chemical processes require a shift from steady‐state operation towards operation under dynamic reaction conditions. This brings along a number of demands for the design of both catalysts and reactors, because it is well‐known that the structure of catalysts is very dynamic. However, in‐depth studies of catalysts and catalytic reactors under such transient conditions have only started recently. This requires studies and advances in the fields of 1) operando spectroscopy including time‐resolved methods, 2) theory with predictive quality, 3) kinetic modelling, 4) design of catalysts by appropriate preparation concepts, and 5) novel/modular reactor designs. An intensive exchange between these scientific disciplines will enable a substantial gain of fundamental knowledge which is urgently required. This concept article highlights recent developments, challenges, and future directions for understanding catalysts under dynamic reaction conditions. PMID:28239429

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fan-Bill Cheung; Joy L. Rempe

2004-06-01

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boilingmore » experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.« less

Transient heat and mass transfer analysis in a porous ceria structure of a novel solar redox reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandran, RB; Bader, R; Lipinski, W

2015-06-01

Thermal transport processes are numerically analyzed for a porous ceria structure undergoing reduction in a novel redox reactor for solar thermochemical fuel production. The cylindrical reactor cavity is formed by an array of annular reactive elements comprising the porous ceria monolith integrated with gas inlet and outlet channels. Two configurations are considered, with the reactor cavity consisting of 10 and 20 reactive elements, respectively. Temperature dependent boundary heat fluxes are obtained on the irradiated cavity wall by solving for the surface radiative exchange using the net radiation method coupled to the heat and mass transfer model of the reactive element.more » Predicted oxygen production rates are in the range 40-60 mu mol s(-1) for the geometries considered. After an initial rise, the average temperature of the reactive element levels off at 1660 and 1680 K for the two geometries, respectively. For the chosen reduction reaction rate model, oxygen release continues after the temperature has leveled off which indicates that the oxygen release reaction is limited by chemical kinetics and/or mass transfer rather than by the heating rate. For a fixed total mass of ceria, the peak oxygen release rate is doubled for the cavity with 20 reactive elements due to lower local oxygen partial pressure. (C) 2015 Elsevier Masson SAS. All rights reserved.« less

CFD simulation of fluid dynamic and biokinetic processes within activated sludge reactors under intermittent aeration regime.

PubMed

Sánchez, F; Rey, H; Viedma, A; Nicolás-Pérez, F; Kaiser, A S; Martínez, M

2018-08-01

Due to the aeration system, biological reactors are the most energy-consuming facilities of convectional WWTPs. Many biological reactors work under intermittent aeration regime; the optimization of the aeration process (air diffuser layout, air flow rate per diffuser, aeration length …) is necessary to ensure an efficient performance; satisfying the effluent requirements with the minimum energy consumption. This work develops a CFD modelling of an activated sludge reactor (ASR) which works under intermittent aeration regime. The model considers the fluid dynamic and biological processes within the ASR. The biological simulation, which is transient, takes into account the intermittent aeration regime. The CFD modelling is employed for the selection of the aeration system of an ASR. Two different aeration configurations are simulated. The model evaluates the aeration power consumption necessary to satisfy the effluent requirements. An improvement of 2.8% in terms of energy consumption is achieved by modifying the air diffuser layout. An analysis of the influence of the air flow rate per diffuser on the ASR performance is carried out. The results show a reduction of 14.5% in the energy consumption of the aeration system when the air flow rate per diffuser is reduced. The model provides an insight into the aeration inefficiencies produced within ASRs. Copyright © 2018 Elsevier Ltd. All rights reserved.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

PHISICS/RELAP5-3D RESULTS FOR EXERCISES II-1 AND II-2 OF THE OECD/NEA MHTGR-350 BENCHMARK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard

2016-03-01

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) High-Temperature Gas-Cooled Reactor (HTGR) Methods group currently leads the Modular High-Temperature Gas-Cooled Reactor (MHTGR) 350 benchmark. The benchmark consists of a set of lattice-depletion, steady-state, and transient problems that can be used by HTGR simulation groups to assess the performance of their code suites. The paper summarizes the results obtained for the first two transient exercises defined for Phase II of the benchmark. The Parallel and Highly Innovative Simulation for INL Code System (PHISICS), coupled with the INL system code RELAP5-3D, was used to generate the results for the Depressurized Conductionmore » Cooldown (DCC) (exercise II-1a) and Pressurized Conduction Cooldown (PCC) (exercise II-2) transients. These exercises require the time-dependent simulation of coupled neutronics and thermal-hydraulics phenomena, and utilize the steady-state solution previously obtained for exercise I-3 of Phase I. This paper also includes a comparison of the benchmark results obtained with a traditional system code “ring” model against a more detailed “block” model that include kinetics feedback on an individual block level and thermal feedbacks on a triangular sub-mesh. The higher spatial fidelity that can be obtained by the block model is illustrated with comparisons of the maximum fuel temperatures, especially in the case of natural convection conditions that dominate the DCC and PCC events. Differences up to 125 K (or 10%) were observed between the ring and block model predictions of the DCC transient, mostly due to the block model’s capability of tracking individual block decay powers and more detailed helium flow distributions. In general, the block model only required DCC and PCC calculation times twice as long as the ring models, and it therefore seems that the additional development and calculation time required for the block model could be worth the gain that can be obtained in the spatial resolution« less

Lithium vapor/aerosol studies. Interim summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Whitlow, G.A.; Bauerle, J.E.; Down, M.G.

1979-04-01

The temperature/cover gas pressure regime, in which detectable lithium aerosol is formed in a static system has been mapped for argon and helium cover gases using a portable He--Ne laser device. At 538/sup 0/C (1000/sup 0/F), lithium aerosol particles were observed over the range 0.5 to 20 torr and 2 to 10 torr for argon and helium respectively. The experimental conditions in this study were more conducive to aerosol formation than in a fusion reactor. In the real reactor system, very high intensity mechanical and thermal disturbances will be made to the liquid lithium. These disturbances, particularly transient increases inmore » lithium vapor pressure appear to be capable of producing high concentrations of optically-dense aerosol. A more detailed study is, therefore, proposed using the basic information generated in these preliminary experiments, as a starting point. Areas recommended include the kinetics of aerosol formation and the occurrence of supersaturated vapor during rapid vapor pressure transients, and also the effect of lithium agitation (falls, jets, splashing, etc.) on aerosol formation.« less

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors. FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

NASA Astrophysics Data System (ADS)

Pauzi, Anas Muhamad; Cioncolini, Andrea; Iacovides, Hector

2015-04-01

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

NASA Astrophysics Data System (ADS)

Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

2017-08-01

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to  <10 MW m-2 while maintaining excellent core confinement, H 98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

Experimental Investigations in a Reactor Cavity Cooling System with Advanced Instrumentation for the Study of Instabilities, Oscillations, and Transients

NASA Astrophysics Data System (ADS)

Tompkins, Casey A.

A research team at University of Wisconsin - Madison designed and constructed a 1/4 height scaled experimental facility to study two-phase natural circulation cooling in a water-based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 28.5 kW m-2 (45 kW) input power, which is equivalent to 14.25 kW m-2 (6.8 MW) at full scale) and pressurized up to 2 bar. The UW-WRCCS facility has been used to study instabilities and oscillations observed during natural circulation flow due to evaporation of the water inventory. During two-phase operation, the system exhibits flow oscillations and excursions, which cause thermal oscillations in the structure. This can cause degradation in the mechanical structure at welds and limit heat transfer to the coolant. The facility is equipped with wire mesh sensors (WMS) that enable high-resolution measurements of the void fraction and steam velocities in order to study the instability's and oscillation's growth and decay during transient operation. Multiple perturbations to the system's operating point in pressure and inlet throttling have shown that the oscillatory behavior present under normal two-phase operating conditions can be damped and removed. Furthermore, with steady-state modeling it was discovered that a flow regime transition instability is the primary cause of oscillations in the UW-WRCCS facility under unperturbed conditions and that proper orifice selection can move the system into a stable operating regime.

Installation of automatic control at experimental breeder reactor II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larson, H.A.; Booty, W.F.; Chick, D.R.

1985-08-01

The Experimental Breeder Reactor II (EBR-II) has been modified to permit automatic control capability. Necessary mechanical and electrical changes were made on a regular control rod position; motor, gears, and controller were replaced. A digital computer system was installed that has the programming capability for varied power profiles. The modifications permit transient testing at EBR-II. Experiments were run that increased power linearly as much as 4 MW/s (16% of initial power of 25 MW(thermal)/s), held power constant, and decreased power at a rate no slower than the increase rate. Thus the performance of the automatic control algorithm, the mechanical andmore » electrical control equipment, and the qualifications of the driver fuel for future power change experiments were all demonstrated.« less

TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.

Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment whilemore » safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional TREAT fuel elements to facilitate the experiment will not inhibit the ability to successfully simulate a RIA for the 2-pin or 3-pin bundle. This new water loop design leaves room for accommodating a larger fuel pin bundle than previously analyzed. The 7-pin fuel bundle in a hexagonal array with similar spacing of fuel pins in a SFR fuel assembly was considered the minimum needed for one central fuel pin to encounter the most correct thermal conditions. The 9-rod fuel bundle in a square array similar in spacing to pins in a LWR fuel assembly would be considered the LWR equivalent. MCNP analysis conducted on a preliminary LWR 9-rod bundle design shows that sufficient energy deposition into the central pin can be achieved well within range to investigate fuel and cladding performance in a simulated RIA. This is achieved by surrounding the flow channel with an additional annulus of water. Findings also show that a highly significant increase in TREAT to specimen power coupling factor (PCF) within the central pin can be achieved by surrounding the experiment with one to two rings of TREAT upgrade fuel assemblies. The experiment design holds promise for the performance evaluation of PWR fuel at extremely high burnup under similar reactor environment conditions.« less

75 FR 38845 - Exelon Generation Company, LLC; Three Mile Island Nuclear Station, Unit No. 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-06

... ignition sources, hot work activities (activities such as welding or grinding), in situ and transient... is \\1/4\\- inch outside diameter tubing used for testing reactor building pressure switches. This... testing on the MU-V-20 backup air supply demonstrated that MU-V- 20 would only stay open for approximately...

CALCINATION AND SINTERING MODELS FOR APPLICATION TO HIGH-TEMPERATURE, SHORT-TIME SULFATION OF CALCIUM-BASED SORBENTS

EPA Science Inventory

To simulate the staged availability of transient high surface area CaO observed in high-temperature flow-reactor data, the rate of calcination of CaCO3 or Ca(OH)2 is described by an empirical modification of the shrinking-core model. The physical model depicts particle decomposi...

TRAC posttest calculations of Semiscale Test S-06-3. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ireland, J.R.; Bleiweis, P.B.

A comparison of Transient Reactor Analysis Code (TRAC) steady-state and transient results with Semiscale Test S-06-3 (US Standard Problem 8) experimental data is discussed. The TRAC model used employs fewer mesh cells than normal data comparison models so that TRAC's ability to obtain reasonable results with less computer time can be assessed. In general, the TRAC results are in good agreement with the data and the major phenomena found in the experiment are reproduced by the code with a substantial reduction in computing times.

ISP33 standard problem on the PACTEL facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Purhonen, H.; Kouhia, J.; Kalli, H.

ISP33 is the first OECD/NEA/CSNI standard problem related to VVER type of pressurized water reactors. The reference reactor of the PACTEL test facility, which was used to carry out the ISP33 experiment, is the VVER-440 reactor, two of which are located near the Finnish city of Loviisa. The objective of the ISP33 test was to study the natural circulation behaviour of VVER-440 reactors at different coolant inventories. Natural circulation was considered as a suitable phenomenon to focus on by the first VVER related ISP due to its importance in most accidents and transients. The behaviour of the natural circulation wasmore » expected to be different compared to Western type of PWRs as a result of the effect of horizontal steam generators and the hot leg loop seals. This ISP was conducted as a blind problem. The experiment was started at full coolant inventory. Single-phase natural circulation transported the energy from the core to the steam generators. The inventory was then reduced stepwise at about 900 s intervals draining 60 kg each time from the bottom of the downcomer. the core power was about 3.7% of the nominal value. The test was terminated after the cladding temperatures began to rise. ATHLET, CATHARE, RELAP5 (MODs 3, 2.5 and 2), RELAP4/MOD6, DINAMIKA and TECH-M4 codes were used in 21 pre- and 20 posttest calculations submitted for the ISP33.« less

Dynamic Modeling and Control Studies of a Two-Stage Bubbling Fluidized Bed Adsorber-Reactor for Solid-Sorbent CO{sub 2} Capture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Modekurti, Srinivasarao; Bhattacharyya, Debangsu; Zitney, Stephen E.

2013-07-31

A one-dimensional, non-isothermal, pressure-driven dynamic model has been developed for a two-stage bubbling fluidized bed (BFB) adsorber-reactor for solid-sorbent carbon dioxide (CO{sub 2}) capture using Aspen Custom Modeler® (ACM). The BFB model for the flow of gas through a continuous phase of downward moving solids considers three regions: emulsion, bubble, and cloud-wake. Both the upper and lower reactor stages are of overflow-type configuration, i.e., the solids leave from the top of each stage. In addition, dynamic models have been developed for the downcomer that transfers solids between the stages and the exit hopper that removes solids from the bottom ofmore » the bed. The models of all auxiliary equipment such as valves and gas distributor have been integrated with the main model of the two-stage adsorber reactor. Using the developed dynamic model, the transient responses of various process variables such as CO{sub 2} capture rate and flue gas outlet temperatures have been studied by simulating typical disturbances such as change in the temperature, flowrate, and composition of the incoming flue gas from pulverized coal-fired power plants. In control studies, the performance of a proportional-integral-derivative (PID) controller, feedback-augmented feedforward controller, and linear model predictive controller (LMPC) are evaluated for maintaining the overall CO{sub 2} capture rate at a desired level in the face of typical disturbances.« less

Independent Qualification of the CIAU Tool Based on the Uncertainty Estimate in the Prediction of Angra 1 NPP Inadvertent Load Rejection Transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borges, Ronaldo C.; D'Auria, Francesco; Alvim, Antonio Carlos M.

2002-07-01

The Code with - the capability of - Internal Assessment of Uncertainty (CIAU) is a tool proposed by the 'Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP)' of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, 'Comissao Nacional de Energia Nuclear', contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in themore » prediction of system code calculations for light water reactors and is internally coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the Angra 1 NPP transient RES-11. 99 originated by an inadvertent complete load rejection that caused the reactor scram when the unit was operating at 99% of nominal power. The current limitation of the 'error' database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated. (authors)« less

Dispersion of a Nanoliter Bolus in Microfluidic Co-Flow.

PubMed

Conway, A J; Saadi, W M; Sinatra, F L; Kowalski, G; Larson, D; Fiering, J

2014-03-01

Microfluidic systems enable reactions and assays on the scale of nanoliters. However, at this scale nonuniformities in sample delivery become significant. To determine the fundamental minimum sample volume required for a particular device, a detailed understanding of mass transport is required. Co-flowing laminar streams are widely used in many devices, but typically only in the steady-state. Because establishing the co-flow steady-state consumes excess sample volume and time, there is a benefit to operating devices in the transient state, which predominates as the volume of the co-flow reactor decreases. Analysis of the co-flow transient has been neglected thus far. In this work we describe the fabrication of a pneumatically controlled microfluidic injector constructed to inject a discrete 50nL bolus into one side of a two-stream co-flow reactor. Using dye for image analysis, injections were performed at a range of flow rates from 0.5-10μL/min, and for comparison we collected the co-flow steady-state data for this range. The results of the image analysis were also compared against theory and simulations for device validation. For evaluation, we established a metric that indicates how well the mass distribution in the bolus injection approximates steady-state co-flow. Using such analysis, transient-state injections can approximate steady-state conditions within predefined errors, allowing straight forward measurements to be performed with reduced reagent consumption.

Bayesian network representing system dynamics in risk analysis of nuclear systems

NASA Astrophysics Data System (ADS)

Varuttamaseni, Athi

2011-12-01

A dynamic Bayesian network (DBN) model is used in conjunction with the alternating conditional expectation (ACE) regression method to analyze the risk associated with the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed operation in the Zion-1 nuclear power plant. The use of the DBN allows the joint probability distribution to be factorized, enabling the analysis to be done on many simpler network structures rather than on one complicated structure. The construction of the DBN model assumes conditional independence relations among certain key reactor parameters. The choice of parameter to model is based on considerations of the macroscopic balance statements governing the behavior of the reactor under a quasi-static assumption. The DBN is used to relate the peak clad temperature to a set of independent variables that are known to be important in determining the success of the feed and bleed operation. A simple linear relationship is then used to relate the clad temperature to the core damage probability. To obtain a quantitative relationship among different nodes in the DBN, surrogates of the RELAP5 reactor transient analysis code are used. These surrogates are generated by applying the ACE algorithm to output data obtained from about 50 RELAP5 cases covering a wide range of the selected independent variables. These surrogates allow important safety parameters such as the fuel clad temperature to be expressed as a function of key reactor parameters such as the coolant temperature and pressure together with important independent variables such as the scram delay time. The time-dependent core damage probability is calculated by sampling the independent variables from their probability distributions and propagate the information up through the Bayesian network to give the clad temperature. With the knowledge of the clad temperature and the assumption that the core damage probability has a one-to-one relationship to it, we have calculated the core damage probably as a function of transient time. The use of the DBN model in combination with ACE allows risk analysis to be performed with much less effort than if the analysis were done using the standard techniques.

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

The effect of carbon crystal structure on treat reactor physics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swanson, R.W.; Harrison, L.J.

1988-01-01

The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbonmore » cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.« less

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

Fast reactor safety and related physics. Volume IV. Phenomenology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-01-01

Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

Influence of High Speed Repetition of Pulsed Streamer Discharge Produced by Polarity-Reversed Traveling Wave on NO Oxidation

NASA Astrophysics Data System (ADS)

Matsuda, Eiji; Kadowaki, Kazunori; Nishimoto, Sakae; Kitani, Isamu

This paper describes experimental results of NO removal from a simulated exhausted gas using a barrier-type plasma reactor subjected to reciprocal traveling wave voltage pulses. A pulse-forming cable was charged and then grounded at one end without any resistance, so that a traveling wave reciprocated along the cable with a change in its polarity because the traveling wave was negatively reflected at the grounded-end. Transient discharge light between point-plane electrodes with a glass barrier was observed using a gated image-intensifier. Photographs of the transient discharge light indicated that many streamer channels extended widely in the gap at the initial stage in the voltage oscillation, while only an intense discharge channel was observed at the latter stage. NO removal tests were carried out using the reciprocal pulse generator and a coaxial plasma reactor with a cylindrical glass-barrier. Results indicated that the discharges at the first and the second polarity reversals contributed largely to the oxidation reaction from NO into NO2, whereas the contribution of the subsequent discharges in the latter stage to NO removal was small.

A time-dependent neutron transport method of characteristics formulation with time derivative propagation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoffman, Adam J., E-mail: adamhoff@umich.edu; Lee, John C., E-mail: jcl@umich.edu

2016-02-15

A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Sourcemore » Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.« less

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

Numerical model of spray combustion in a single cylinder diesel engine

NASA Astrophysics Data System (ADS)

Acampora, Luigi; Sequino, Luigi; Nigro, Giancarlo; Continillo, Gaetano; Vaglieco, Bianca Maria

2017-11-01

A numerical model is developed for predicting the pressure cycle from Intake Valve Closing (IVC) to the Exhaust Valve Opening (EVO) events. The model is based on a modified one-dimensional (1D) Musculus and Kattke spray model, coupled with a zero-dimensional (0D) non-adiabatic transient Fed-Batch reactor model. The 1D spray model provides an estimate of the fuel evaporation rate during the injection phenomenon, as a function of time. The 0D Fed-Batch reactor model describes combustion. The main goal of adopting a 0D (perfectly stirred) model is to use highly detailed reaction mechanisms for Diesel fuel combustion in air, while keeping the computational cost as low as possible. The proposed model is validated by comparing its predictions with experimental data of pressure obtained from an optical single cylinder Diesel engine.

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Development of advanced high heat flux and plasma-facing materials

NASA Astrophysics Data System (ADS)

Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

2017-09-01

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Startup of RAPID-L Lunar Base Reactor by Lithium Release Module

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept RAPID-L to be combined with thermoelectric power conversion system for lunar base power system is demonstrated. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of conventional B4C rods or Be reflectors. These systems are effective independent of the magnitude and direction of the gravity force. In 2006, however, the following design amendment has been made. 1) B4C poison rods were added to ensure criticality safety in unintended positive reactivity insertion by LRMs due to fire in the launch phase accident; because LRM freeze seal melts at 800°C which result in positive reactivity insertion. 2) Lower hot standby temperature of 200°C was adopted instead of conventional 800°C to reduce the external power at the startup. In this paper, development of the LRM orifice which dominates the startup transient of RAPID-L is discussed. An attention was focused how to achieve sufficiently small flow rate of 6Li in the orifice because it enables moderate positive reactivity insertion rate. The LRM orifice performance has been confirmed using 0.5 mm diameter SUS316 orifice/lithium flow test setup in the glove box.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my; Cioncolini, Andrea; Iacovides, Hector

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software calledmore » FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.« less

Modeling Hybrid Nuclear Systems With Chilled-Water Storage

DOE PAGES

Misenheimer, Corey T.; Terry, Stephen D.

2016-06-27

Air-conditioning loads during the warmer months of the year are large contributors to an increase in the daily peak electrical demand. Traditionally, utility companies boost output to meet daily cooling load spikes, often using expensive and polluting fossil fuel plants to match the demand. Likewise, heating, ventilation, and air conditioning (HVAC) system components must be sized to meet these peak cooling loads. However, the use of a properly sized stratified chilled-water storage system in conjunction with conventional HVAC system components can shift daily energy peaks from cooling loads to off-peak hours. This process is examined in light of the recentmore » development of small modular nuclear reactors (SMRs). In this paper, primary components of an air-conditioning system with a stratified chilled-water storage tank were modeled in FORTRAN 95. A basic chiller operation criterion was employed. Simulation results confirmed earlier work that the air-conditioning system with thermal energy storage (TES) capabilities not only reduced daily peaks in energy demand due to facility cooling loads but also shifted the energy demand from on-peak to off-peak hours, thereby creating a more flattened total electricity demand profile. Thus, coupling chilled-water storage-supplemented HVAC systems to SMRs is appealing because of the decrease in necessary reactor power cycling, and subsequently reduced associated thermal stresses in reactor system materials, to meet daily fluctuations in cooling demand. Finally and also, such a system can be used as a thermal sink during reactor transients or a buffer due to renewable intermittency in a nuclear hybrid energy system (NHES).« less

Modeling Hybrid Nuclear Systems With Chilled-Water Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Misenheimer, Corey T.; Terry, Stephen D.

Air-conditioning loads during the warmer months of the year are large contributors to an increase in the daily peak electrical demand. Traditionally, utility companies boost output to meet daily cooling load spikes, often using expensive and polluting fossil fuel plants to match the demand. Likewise, heating, ventilation, and air conditioning (HVAC) system components must be sized to meet these peak cooling loads. However, the use of a properly sized stratified chilled-water storage system in conjunction with conventional HVAC system components can shift daily energy peaks from cooling loads to off-peak hours. This process is examined in light of the recentmore » development of small modular nuclear reactors (SMRs). In this paper, primary components of an air-conditioning system with a stratified chilled-water storage tank were modeled in FORTRAN 95. A basic chiller operation criterion was employed. Simulation results confirmed earlier work that the air-conditioning system with thermal energy storage (TES) capabilities not only reduced daily peaks in energy demand due to facility cooling loads but also shifted the energy demand from on-peak to off-peak hours, thereby creating a more flattened total electricity demand profile. Thus, coupling chilled-water storage-supplemented HVAC systems to SMRs is appealing because of the decrease in necessary reactor power cycling, and subsequently reduced associated thermal stresses in reactor system materials, to meet daily fluctuations in cooling demand. Finally and also, such a system can be used as a thermal sink during reactor transients or a buffer due to renewable intermittency in a nuclear hybrid energy system (NHES).« less

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

NASA Astrophysics Data System (ADS)

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Sugawara, Takanori; Kurata, Yuji; Takei, Hayanori; Saito, Shigeru; Sasa, Toshinobu; Obayashi, Hironari

2011-08-01

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the "double-strata" fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolstad, J.W.; Haarman, R.A.

The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

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.

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.

Thermal-hydraulic modeling needs for passive reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

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,more » 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.« less

Nuclear reactor neutron shielding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

Process and apparatus for adding and removing particles from pressurized reactors

DOEpatents

Milligan, John D.

1983-01-01

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Effects of imperfect mixing on low-density polyethylene reactor dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Villa, C.M.; Dihora, J.O.; Ray, W.H.

1998-07-01

Earlier work considered the effect of feed conditions and controller configuration on the runaway behavior of LDPE autoclave reactors assuming a perfectly mixed reactor. This study provides additional insight on the dynamics of such reactors by using an imperfectly mixed reactor model and bifurcation analysis to show the changes in the stability region when there is imperfect macroscale mixing. The presence of imperfect mixing substantially increases the range of stable operation of the reactor and makes the process much easier to control than for a perfectly mixed reactor. The results of model analysis and simulations are used to identify somemore » of the conditions that lead to unstable reactor behavior and to suggest ways to avoid reactor runaway or reactor extinction during grade transitions and other process operation disturbances.« less

The role of accelerators in the nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, Hiroshi.

1990-01-01

The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the usemore » of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.« less

Reactivity effects of moderator voids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahlfeld, C.E.; Pryor, R.J.

1975-01-01

Reactivity worths for large moderator voids similar to those produced by steaming in postulated reactor transients were measured in the Process Development Pile (PDP) reactor. The experimental results were compared to the computed void worths obtained from techniques currently used in routine safety analyses. Neutron energy spectrum measurements were used to verify a modified lattice pattern that correctly computed the measured spectrum, and consequently, improved macroscopic cross sections. In addition, a special two-dimensional transport calculation was performed to obtain an axially defined diffusion coefficient for the void region. The combination of the modified lattice calculations and the axial diffusion coefficientmore » yielded void reactivity worths which agreed very well with experiment. It was concluded that the computational modules available in the JOSHUA system (GLASS, GRIMHX) would yield accurate void reactivity worths in SLR--SRP safety analysis studies, provided the above mentioned modifications were made.« less

Safety monitoring and reactor transient interpreter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hench, J. E.; Fukushima, T. Y.

1983-12-20

An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18,20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24)more » for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18,20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.« less

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Monitoring pH and electric conductivity in an EBPR sequencing batch reactor.

PubMed

Serralta, J; Borrás, L; Blanco, C; Barat, R; Seco, A

2004-01-01

This paper presents laboratory-scale experimentation carried out to study enhanced biological phosphorus removal. Two anaerobic aerobic (A/O) sequencing batch reactors (SBR) have been operated during more than one year to investigate the information provided by monitoring pH and electric conductivity under stationary and transient conditions. Continuous measurements of these parameters allow detecting the end of anaerobic phosphorus release, of aerobic phosphorus uptake and of initial denitrification, as well as incomplete acetic acid uptake. These results suggest the possibility of using pH and electric conductivity as control parameters to determine the length of both anaerobic and aerobic phases in an A/O SBR. More valuable information provided by monitoring pH and electric conductivity is the relation between the amount of phosphorus released and the conductivity increase observed during the anaerobic stages and which group of bacteria (heterotrophic or polyphosphate accumulating) is carrying out the denitrification process.

Safety and control of accelerator-driven subcritical systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rief, H.; Takahashi, H.

1995-10-01

To study control and safety of accelertor driven nuclear systems, a one point kinetic model was developed and programed. It deals with fast transients as a function of reactivity insertion. Doppler feedback, and the intensity of an external neutron source. The model allows for a simultaneous calculation of an equivalent critical reactor. It was validated by a comparison with a benchmark specified by the Nuclear Energy Agency Committee of Reactor Physics. Additional features are the possibility of inserting a linear or quadratic time dependent reactivity ramp which may account for gravity induced accidents like earthquakes, the possibility to shut downmore » the external neutron source by an exponential decay law of the form exp({minus}t/{tau}), and a graphical display of the power and reactivity changes. The calculations revealed that such boosters behave quite benignly even if they are only slightly subcritical.« less

Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, Javier; Baker, Benjamin Allen; Schunert, Sebastian

The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition,more » this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.« less

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

A cermet fuel reactor for nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Kruger, Gordon

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

In situ monitored in-pile creep testing of zirconium alloys

NASA Astrophysics Data System (ADS)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

Fuel loading of PeBR for a long operation life on the lunar surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schriener, T. M.; Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM; El-Genk, M. S.

2012-07-01

The Pellet Bed Reactor (PeBR) power system could provide 99.3 kW e to a lunar outpost for 66 full power years and is designed for no single point failures. The core of this fast energy spectrum reactor consists of three sectors that are neutronically and thermally coupled, but hydraulically independent. Each sector has a separate Closed Brayton Cycle (CBC) loop for energy conversion and separate water heat-pipes radiator panels for heat rejection. He-Xe (40 g/mole) binary gas mixture serves as the reactor coolant and CBC working fluid. On the lunar surface, the emplaced PeBR below grade is loaded with sphericalmore » fuel pellets (1-cm in dia.). It is launched unfueled and the pellets are launched in separate subcritical canisters, one for each core sector. This paper numerically simulates the transient loading of a core sector with fuel pellets on the Moon. The simulation accounts for the dynamic interaction of the pellets during loading and calculates the axial and radial distributions of the volume porosity in the sector. The pellets pack randomly with a volume porosity of 0.39 - 0.41 throughout most of the sector, except near the walls the local porosity is higher. (authors)« less

Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

DOE PAGES

Valentin, Francisco I.; Artoun, Narbeh; Anderson, Ryan; ...

2016-12-01

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical resultsmore » were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.« less

Low nitrous oxide production through nitrifier-denitrification in intermittent-feed high-rate nitritation reactors.

PubMed

Su, Qingxian; Ma, Chun; Domingo-Félez, Carlos; Kiil, Anne Sofie; Thamdrup, Bo; Jensen, Marlene Mark; Smets, Barth F

2017-10-15

Nitrous oxide (N 2 O) production from autotrophic nitrogen conversion processes, especially nitritation systems, can be significant, requires understanding and calls for mitigation. In this study, the rates and pathways of N 2 O production were quantified in two lab-scale sequencing batch reactors operated with intermittent feeding and demonstrating long-term and high-rate nitritation. The resulting reactor biomass was highly enriched in ammonia-oxidizing bacteria, and converted ∼93 ± 14% of the oxidized ammonium to nitrite. The low DO set-point combined with intermittent feeding was sufficient to maintain high nitritation efficiency and high nitritation rates at 20-26 °C over a period of ∼300 days. Even at the high nitritation efficiencies, net N 2 O production was low (∼2% of the oxidized ammonium). Net N 2 O production rates transiently increased with a rise in pH after each feeding, suggesting a potential effect of pH on N 2 O production. In situ application of 15 N labeled substrates revealed nitrifier denitrification as the dominant pathway of N 2 O production. Our study highlights operational conditions that minimize N 2 O emission from two-stage autotrophic nitrogen removal systems. Copyright © 2017 Elsevier Ltd. All rights reserved.

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

Baseline Concept Description of a Small Modular High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans D.

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

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.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 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 required...

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... 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 required...

KINETICS OF TREAT USED AS A TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, Douglas E.; Orr, Richard

1993-01-01

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, D.E.; Orr, R.

1993-12-07

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Summary of NR Program Prometheus Efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Control Means for Reactor

DOEpatents

Manley, J. H.

1961-06-27

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

Low temperature pre-treatment of domestic sewage in an anaerobic hybrid or an anaerobic filter reactor.

PubMed

Elmitwalli, Tarek A; Sklyar, Vladimir; Zeeman, Grietje; Lettinga, Gatze

2002-05-01

The pre-treatment of domestic sewage for removal of suspended solids (SS) at a process temperature of 13 degrees C and an hydraulic retention time (HRT) of 4 h was investigated in an anaerobic filter (AF) and anaerobic hybrid (AH) reactor. The AF and the top of the AH reactor consisted of vertical sheets of reticulated polyurethane foam (RPF) with knobs. All biomass in the AF was only in attached form to avoid clogging and sludge washout. The AF reactor showed a significantly higher removal of total and suspended chemical oxygen demand (COD) than the AH reactor, respectively, 55% and 82% in the AF reactor and 34% and 53% in the AH reactor. Because the reactors were operated at a short HRT and low temperature, the hydrolysis, acidification and methanogenesis based on the influent COD were limited to, respectively, 12%, 21% and 23% for the AF reactor and 12%, 17% and 16% for the AH reactor. The excess sludge from the AH reactor was more stabilised and had a better settling capacity and dewaterability. However, the excess sludge from both the AH and AF reactors needed stabilisation. Therefore, the AF reactor is recommended for the pretreatment of domestic sewage at low temperatures.

Nuclear reactor cavity floor passive heat removal system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Methods and apparatuses for deoxygenating pyrolysis oil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baird, Lance Awender; Brandvold, Timothy A.; Frey, Stanley Joseph

Methods and apparatuses are provided for deoxygenating pyrolysis oil. A method includes contacting a pyrolysis oil with a deoxygenation catalyst in a first reactor at deoxygenation conditions to produce a first reactor effluent. The first reactor effluent has a first oxygen concentration and a first hydrogen concentration, based on hydrocarbons in the first reactor effluent, and the first reactor effluent includes an aromatic compound. The first reactor effluent is contacted with a dehydrogenation catalyst in a second reactor at conditions that deoxygenate the first reactor effluent while preserving the aromatic compound to produce a second reactor effluent. The second reactormore » effluent has a second oxygen concentration lower than the first oxygen concentration and a second hydrogen concentration that is equal to or lower than the first hydrogen concentration, where the second oxygen concentration and the second hydrogen concentration are based on the hydrocarbons in the second reactor effluent.« less

SNPSAM - Space Nuclear Power System Analysis Model

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Seo, Jong T.

The current version of SNPSAM is described, and the results of the integrated thermoeletric SP-100 system performance studies using SNPSAM are reported. The electric power output, conversion efficiency, coolant temperatures, and specific pumping power of the system are calculated as functions of the reactor thermal power and the liquid metal coolant type (Li or NaK-78) during steady state operation. The transient behavior of the system is also discussed.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

DOE Office of Scientific and Technical Information (OSTI.GOV)

farahani, A.A.; Corradini, M.L.

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for themore » foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.« less

COMPUTATIONAL MODELING OF CIRCULATING FLUIDIZED BED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ibrahim, Essam A

2013-01-09

Details of numerical simulations of two-phase gas-solid turbulent flow in the riser section of Circulating Fluidized Bed Reactor (CFBR) using Computational Fluid Dynamics (CFD) technique are reported. Two CFBR riser configurations are considered and modeled. Each of these two riser models consist of inlet, exit, connecting elbows and a main pipe. Both riser configurations are cylindrical and have the same diameter but differ in their inlet lengths and main pipe height to enable investigation of riser geometrical scaling effects. In addition, two types of solid particles are exploited in the solid phase of the two-phase gas-solid riser flow simulations tomore » study the influence of solid loading ratio on flow patterns. The gaseous phase in the two-phase flow is represented by standard atmospheric air. The CFD-based FLUENT software is employed to obtain steady state and transient solutions for flow modulations in the riser. The physical dimensions, types and numbers of computation meshes, and solution methodology utilized in the present work are stated. Flow parameters, such as static and dynamic pressure, species velocity, and volume fractions are monitored and analyzed. The differences in the computational results between the two models, under steady and transient conditions, are compared, contrasted, and discussed.« less

Modeling of local steam condensation on walls in presence of non-condensable gases. Application to a loca calculation in reactor containment using the multidimensional geyser/tonus code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benet, L.V.; Caroli, C.; Cornet, P.

1995-09-01

This paper reports part of a study of possible severe pressurized water reactor (PWR) accidents. The need for containment modeling, and in particular for a hydrogen risk study, was reinforced in France after 1990, with the requirement that severe accidents must be taken into account in the design of future plants. This new need of assessing the transient local hydrogen concentration led to the development, in the Mechanical Engineering and Technology Department of the French Atomic Energy Commission (CEA/DMT), of the multidimensional code GEYSER/TONUS for containment analysis. A detailed example of the use of this code is presented. The mixturemore » consisted of noncondensable gases (air or air plus hydrogen) and water vapor and liquid water. This is described by a compressible homogeneous two-phase flow model and wall condensation is based on the Chilton-Colburn formula and the analogy between heat and mass transfer. Results are given for a transient two-dimensional axially-symmetric computation for the first hour of a simplified accident sequence. In this there was an initial injection of a large amount of water vapor followed by a smaller amount and by hydrogen injection.« less

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2008-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-12-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Westinghouse Small Modular Reactor balance of plant and supporting systems design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Stansbury, C.; Taylor, C.

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 second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operationmore » of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)« less

A physical description of fission product behavior fuels for advanced power reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

Non-Nuclear Validation Test Results of a Closed Brayton Cycle Test-Loop

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

Both NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, or for next generation nuclear power plants on earth. Although open Brayton cycles are in use for many applications (combined cycle power plants, aircraft engines), only a few closed Brayton cycles have been tested. Experience with closed Brayton cycles coupled to nuclear reactors is even more limited and current projections of Brayton cycle performance are based on analytic models. This report describes and compares experimental results with model predictions from a series of non-nuclear tests using a small scale closed loop Brayton cycle available at Sandia National Laboratories. A substantial amount of testing has been performed, and the information is being used to help validate models. In this report we summarize the results from three kinds of tests. These tests include: 1) test results that are useful for validating the characteristic flow curves of the turbomachinery for various gases ranging from ideal gases (Ar or Ar/He) to non-ideal gases such as CO2, 2) test results that represent shut down transients and decay heat removal capability of Brayton loops after reactor shut down, and 3) tests that map a range of operating power versus shaft speed curve and turbine inlet temperature that are useful for predicting stable operating conditions during both normal and off-normal operating behavior. These tests reveal significant interactions between the reactor and balance of plant. Specifically these results predict limited speed up behavior of the turbomachinery caused by loss of load, the conditions for stable operation, and for direct cooled reactors, the tests reveal that the coast down behavior during loss of power events can extend for hours provided the ultimate heat sink remains available.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R J; Gottula, R C; Holcomb, E E

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

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.

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

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.

The role of nuclear reactors in space exploration and development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« less

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

Unmixed fuel processors and methods for using the same

DOEpatents

Kulkarni, Parag Prakash; Cui, Zhe

2010-08-24

Disclosed herein are unmixed fuel processors and methods for using the same. In one embodiment, an unmixed fuel processor comprises: an oxidation reactor comprising an oxidation portion and a gasifier, a CO.sub.2 acceptor reactor, and a regeneration reactor. The oxidation portion comprises an air inlet, effluent outlet, and an oxygen transfer material. The gasifier comprises a solid hydrocarbon fuel inlet, a solids outlet, and a syngas outlet. The CO.sub.2 acceptor reactor comprises a water inlet, a hydrogen outlet, and a CO.sub.2 sorbent, and is configured to receive syngas from the gasifier. The regeneration reactor comprises a water inlet and a CO.sub.2 stream outlet. The regeneration reactor is configured to receive spent CO.sub.2 adsorption material from the gasification reactor and to return regenerated CO.sub.2 adsorption material to the gasification reactor, and configured to receive oxidized oxygen transfer material from the oxidation reactor and to return reduced oxygen transfer material to the oxidation reactor.

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.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

DOEpatents

West, J.M.; Weills, J.T.

1960-03-15

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

Thermal-hydraulic interfacing code modules for CANDU reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients

NASA Astrophysics Data System (ADS)

Gallais, L.; Burla, R.; Martin, F.; Richaud, J. C.; Volle, G.; Pontillon, M.; Capdevila, H.; Pontillon, Y.

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients.

PubMed

Gallais, L; Burla, R; Martin, F; Richaud, J C; Volle, G; Pontillon, M; Capdevila, H; Pontillon, Y

2018-01-01

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

76 FR 55718 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor...'' for reactor coolant system (RCS) components, as mentioned in 10 CFR 50 Appendix A, GDC-4. The...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

75 FR 58449 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-24

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor... would result in a major inconvenience. Dated: September 17, 2010. Antonio Dias, Chief, Reactor Safety...

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

Function of university reactors in operator licensing training for nuclear utilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

Numerical Simulations of a 96-rod Polysilicon CVD Reactor

NASA Astrophysics Data System (ADS)

Guoqiang, Tang; Cong, Chen; Yifang, Cai; Bing, Zong; Yanguo, Cai; Tihu, Wang

2018-05-01

With the rapid development of the photovoltaic industry, pressurized Siemens belljar-type polysilicon CVD reactors have been enlarged from 24 rods to 96 rods in less than 10 years aimed at much greater single-reactor productivity. A CFD model of an industry-scale 96-rod CVD reactor was established to study the internal temperature distribution and the flow field of the reactor. Numerical simulations were carried out and compared with actual growth results from a real CVD reactor. Factors affecting polysilicon depositions such as inlet gas injections, flow field, and temperature distribution in the CVD reactor are studied.

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.

Characteristics and Dose Levels for Spent Reactor Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Reactor operation environmental information document

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lister, Tedd E; Parkman, Jacob A; Diaz Aldana, Luis A

A method of recovering metals from electronic waste comprises providing a powder comprising electronic waste in at least a first reactor and a second reactor and providing an electrolyte comprising at least ferric ions in an electrochemical cell in fluid communication with the first reactor and the second reactor. The method further includes contacting the powders within the first reactor and the second reactor with the electrolyte to dissolve at least one base metal from each reactor into the electrolyte and reduce at least some of the ferric ions to ferrous ions. The ferrous ions are oxidized at an anodemore » of the electrochemical cell to regenerate the ferric ions. The powder within the second reactor comprises a higher weight percent of the at least one base metal than the powder in the first reactor. Additional methods of recovering metals from electronic waste are also described, as well as an apparatus of recovering metals from electronic waste.« less

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Experiences in utilization of research reactors in Yugoslavia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevanovic, Vladimir D.; Stosic, Zoran V.; Kiera, Michael

2002-07-01

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give completemore » insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes. (authors)« less

Two-phase flow in the cooling circuit of a cryogenic rocket engine

NASA Astrophysics Data System (ADS)

Preclik, D.

1992-07-01

Transient two-phase flow was investigated for the hydrogen cooling circuit of the HM7 rocket engine. The nuclear reactor code ATHLET/THESEUS was adapted to cryogenics and applied to both principal and prototype experiments for validation and simulation purposes. The cooling circuit two-phase flow simulation focused on the hydrogen prechilling and pump transient phase prior to ignition. Both a single- and a multichannel model were designed and employed for a valve leakage flow, a nominal prechilling flow, and a prechilling with a subsequent pump-transient flow. The latter case was performed in order to evaluate the difference between a nominal and a delayed turbo-pump start-up. It was found that an extension of the nominal prechilling sequence in the order of 1 second is sufficient to finally provide for liquid injection conditions of hydrogen which, as commonly known, is undesirable for smooth ignition and engine starting transients.

Inter-Disciplinary Collaboration in Support of the Post-Standby TREAT Mission

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark; Baker, Benjamin; Ortensi, Javier

Although analysis methods have advanced significantly in the last two decades, high fidelity multi- physics methods for reactors systems have been under development for only a few years and are not presently mature nor deployed. Furthermore, very few methods provide the ability to simulate rapid transients in three dimensions. Data for validation of advanced time-dependent multi- physics is sparse; at TREAT, historical data were not collected for the purpose of validating three-dimensional methods, let alone multi-physics simulations. Existing data continues to be collected to attempt to simulate the behavior of experiments and calibration transients, but it will be insufficient formore » the complete validation of analysis methods used for TREAT transient simulations. Hence, a 2018 restart will most likely occur without the direct application of advanced modeling and simulation methods. At present, the current INL modeling and simulation team plans to work with TREAT operations staff in performing reactor simulations with MAMMOTH, in parallel with the software packages currently being used in preparation for core restart (e.g., MCNP5, RELAP5, ABAQUS). The TREAT team has also requested specific measurements to be performed during startup testing, currently scheduled to run from February to August of 2018. These startup measurements will be crucial in validating the new analysis methods in preparation for ultimate application for TREAT operations and experiment design. This document describes the collaboration between modeling and simulation staff and restart, operations, instrumentation and experiment development teams to be able to effectively interact and achieve successful validation work during restart testing.« less

TECHNICAL REVIEW OF ZPR-I ACCIDENTAL TRANSIENT--THE POWER EXCURSION, EXPOSURES, AND CLINICAL DATA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brittan, R. O.; Hasterlik, R. J.; Marinelli, L. D.

1953-01-26

On June 2. 1952. a large reactivity change was made manually in a ZPR-1 assembly causing a power excursion of about one kwh, which resulted in damage to the reactor core components and radiation exposure of some of the operating personnel to perhaps several hundred rep. A description is presented of the incident, estimates are made of the exposures, and early clinical data are summarized. (C.H.)

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

Real time monitoring of environmental crack growth in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, D.; Diehl, C.G.

1988-01-01

A comprehensive field test program was recently completed at several Boiling Water Reactors (BWR) to quantify the effect of coolant impurities on the initiation and growth of stress corrosion cracks. A new technology was utilized which allows for real time monitoring of stress corrosion crack growth rates. The BWR environments were characterized using Ion Chromatography and Electro Chemical Potential (ECP) measurements. The effects of typical water chemistry transients and startups were quantified.

Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH) Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

he Liquid Phase Methanol (LPMEOW) Demonstration Project at Kingsport Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership) to produce methanol from coal-derived synthesis gas (syngas). Air Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOEP Process Demonstration Unit was built at a site located at the Eastman coal-to-chemicals complex in Kingsport. The LPMEOHW Demonstration Facility completed its first year of operation on 02 April 1998. The LPMEOW Demonstration Facility also completed themore » longest continuous operating run (65 days) on 21 April 1998. Catalyst activity, as defined by the ratio of the rate constant at any point in time to the rate constant for freshly reduced catalyst (as determined in the laboratory autoclave), was monitored throughout the reporting period. During a six-week test at a reactor temperature of 225oC and Balanced Gas flowrate of 700 KSCFH, the rate of decline in catalyst activity was steady at 0.29-0.36% per day. During a second one-month test at a reactor temperature of 220oC and a Balanced Gas flowrate of 550-600 KSCFH, the rate of decline in catalyst activity was 0.4% per day, which matched the pefiorrnance at 225"C, as well as the 4-month proof-of-concept run at the LaPorte AFDU in 1988/89. Beginning on 08 May 1998, the LPMEOW Reactor temperature was increased to 235oC, which was the operating temperature tier the December 1997 restart with the fresh charge of catalyst (50'Yo of design loading). The flowrate of the primary syngas feed stream (Balanced Gas) was also increased to 700-750 KSCFH. During two stable operating periods between 08 May and 09 June 1998, the average catalyst deactivation rate was 0.8% per day. Due to the scatter of the statistical analysis of the results, this test was extended to better quanti& the catalyst aging behavior. During the reporting perio~ two batches of fresh catalyst were activated and transferred to the reactor (on 02 April and 20 June 1998). The weight of catalyst in the LPMEOW Reactor has reached 80% of the design value. At the end of the reporting period, a step-change in the pressure-drop profile within the LPMEOW Reactor and an increase in the pressure of the steam system which provides cooling to the LPMEOW Reactor were observed. No change in the calculated activity of the catalyst was detected during either of these transients. These parameters will be monitored closely for any additional changes.« less

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

System and method for temperature control in an oxygen transport membrane based reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

Modification of UASB reactor by using CFD simulations for enhanced treatment of municipal sewage.

PubMed

Das, Suprotim; Sarkar, Supriya; Chaudhari, Sanjeev

2018-02-01

Up-flow anaerobic sludge blanket (UASB) has been in use since last few decades for the treatment of organic wastewaters. However, the performance of UASB reactor is quite low for treatment of low strength wastewaters (LSWs) due to less biogas production leading to poor mixing. In the present research work, a modification was done in the design of UASB to improve mixing of reactor liquid which is important to enhance the reactor performance. The modified UASB (MUASB) reactor was designed by providing a slanted baffle along the height of the reactor having an angle of 5.7° with the vertical wall. A two-dimensional computational fluid dynamics (CFD) simulation of three phase gas-liquid-solid flow in MUASB reactor was performed and compared with conventional UASB reactor. The CFD study indicated better mixing in terms of vorticity magnitude in MUASB reactor as compared to conventional UASB, which was reflected in the reactor performance. The performance of MUASB was compared with conventional UASB reactor for the onsite treatment of domestic sewage as LSW. Around 16% higher total chemical oxygen demand removal efficiency was observed in MUASB reactor as compared to conventional UASB during this study. Therefore, this MUASB model demonstrates a qualitative relationship between mixing and performance during the treatment of LSW. From the study, it seems that MUASB holds promise for field applications.

Phased Development of Accident Tolerant Fue

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, Shannon M.; Carmack, W. Jon

2016-09-01

The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modelingmore » predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology readiness levels (TRL) defined for nuclear fuel development. TRL 1–3 corresponds to the “proof-of-concept” stage (Phase 1), TRL 4–6 to “proof-of-principle” (Phase 2), and TRL 7–9 to “proof-of-performance” (Phase 3). This paper will provide an overview of the anticipated activities within each phase of development and will provide an update on the current ATF development status.« less

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with the Advanced TREAT loop, and the stress placed on these calculations reflects this. Additional material on performance and safety is provided. Backup calculations on calculations of plutonium-release limits are described. Cost and schedule information for the Upgrade are presented.« less

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-08

... Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and... 20, 2012 (77 FR 42771), ``License Renewal for the Dow Chemical TRIGA Research Reactor,'' to inform... Chemical Company which would authorize continued operation of the Dow TRIGA Research Reactor. The notice...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.