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Research in the field of passive system reliability analysis is garnering sharp interest in the nuclear community. Passive systems are being utilized extensively in current- and future-generation reactors for their normal operations as well as for safety critical operations during any accidental conditions. In this paper, we present a methodology called Analysis of Passive System ReliAbility Plus (APSRA+) for evaluating reliability of passive systems. This methodology is an improved version of the existing APSRA methodology. The methodology has been applied to the passive isolation condenser system (ICS) of the AHWR (Advanced Heavy Water Reactor). With the help of the APSRA+ methodology, the probability of the passive ICS failing to maintain the clad temperature under 400°C is estimated to be of the order 1?10-10. Important features of APSRA+ are the following. First, it provides an integrated dynamic reliability method for the consistent treatment of dynamic failure characteristics such as multistate failure, fault increment, and time-dependent failure rate of components of passive systems. Second, this methodology overcomes the issue of process parameter treatment by just the probability density function or by root cause analysis, by segregating the parameters into dependent and independent process parameters and then giving a proper treatment to each of them separately. Third, the methodology treats the model uncertainties and independent process parameter variations in a consistent manner. In APSRA+, the important parameters affecting the passive system under consideration are identified using sensitivity analysis. To evaluate the system performance, a best-estimate system code is used with due consideration of the uncertainties in empirical models. A failure surface is generated by varying all the identified important parameters; variation from the nominal values of these parameters affects the system performance significantly. These parameters are then segregated into dependent and independent categories. For dependent parameters, it is attributed that the variations of process parameters are mainly due to malfunction of mechanical components or control systems, and hence, root cause analysis is performed. The probability of these dependent parameter variations is estimated using a dynamic reliability methodology based on Monte Carlo simulation. The dynamic failure characteristics of the identified causal component/system are accounted for in calculating these probabilities. For the treatment of independent process parameters, using APSRA+ suggests adopting and integrating classical data-fitting techniques or mathematical models. In the next steps, a response surface-based metamodel is formulated using the generated failure points. The probability of the system being in the failure zone is estimated by sampling and analyzing a sufficiently large number of samples for all the dependent and independent process parameters based on the probability of variations of these parameters, which were estimated using dynamic reliability methodology. [ABSTRACT FROM AUTHOR]

The Missouri University of Science and Technology Subcritical Assembly has been brought back into service and upgraded with a new neutron detection system and Internet accessibility. Before the upgrade, neutron counting was possible in only one location. Using a movable detection system housed in acrylic tubes, measurements can now be taken in any empty fuel location and at any height within the tube, making three-dimensional flux mapping possible. By connecting the new detection system to a Canberra Lynx Digital Signal Analyzer, remote users can have limited data-collecting capabilities. To further enhance the potential of the facility, a Monte Carlo N-Particle transport code (MCNP) model of the subcritical assembly was created and validated by comparing its simulated predictions to experiments conducted at the facility. An approach to the criticality experiment using the 1/M approximation showed that the MCNP model accurately predicts keff if the detectors are placed between 27 and 36 cm from the neutron source. The results of an axial flux measurement experiment taken 20.3 cm from the neutron source differed from the MCNP-simulated results by an average of 12%. Finally, the validated MCNP model was used to show the effect of removing the facility's fixed detector tube and redistributing its fuel. MCNP simulation predicts that the new configuration would increase the multiplication factor from 0.73481 ± 0.00008 to 0.76844 ± 0.00004. [ABSTRACT FROM AUTHOR]

This paper introduces VEGAS, a lightweight and fast-executing fuel cycle simulator primarily intended to augment higher-fidelity fuel cycle simulators. VEGAS can act as a preconditioner as well as a platform for implementing endogenous decision making within these simulators. In preconditioner mode, VEGAS offers an iterative method for accelerating the convergence of fuel cycle optimization problems while ensuring that material balance and other constraints are met. The methodology utilized by VEGAS, particularly its mass balance and reactor deployment algorithms, is presented here. Unique to the VEGAS simulator is a rollback feature that undoes reactor build decisions if material balance constraints are violated. Included in the VEGAS code is an economics package, also documented here, that calculates the evolving levelized cost of electricity. Benchmark comparisons against the VISION (Verifiable Fuel Cycle Simulation Model) simulator are also presented, along with an example of VEGAS's preconditioning capability in which the objective is to achieve a target system transuranics inventory while installing as little reprocessing capacity as possible. [ABSTRACT FROM AUTHOR]

This paper presents the core design and performance characteristics of the Novel Modular Reactor (NMR-50), a 50-MW(electric) small modular reactor. NMR-50 is a boiling water reactor with natural-circulation cooling and two layers of passive safety systems that enable the reactor to withstand prolonged station blackout and loss of ultimate heat sink accidents. The main goal in the core design is to achieve a long-life core (~10 years) without refueling for deployment in remote sites. Through assembly design studies with the CASMO-4 lattice code and coupled neutronics and thermal-hydraulic core analyses with the PARCS and RELAP5 codes, a preliminary NMR-50 core design has been developed to meet the 10-year cycle length with an average fuel enrichment of 4.75 wt% and a maximum enrichment of 5.0 wt%. The calculated fuel temperature coefficient and coolant void coefficient provide adequate negative reactivity feedbacks. The maximum fuel linear power density throughout the 10-year burn cycle is 18.7 kW/m, and the minimum critical power ratio is 2.07, both of which meet the selected design limits with significant margins. Preliminary safety analyses using the RELAP5 code show that the core will remain covered during the entire transient procedure of two design-basis loss-of-coolant accidents. These results indicate that the targeted 10-year cycle length is achievable while satisfying the operation and safety-related design criteria with sufficient margins. [ABSTRACT FROM AUTHOR]

At elevated temperatures, the corrosion of carbon steel and Type 304 stainless steel is high in media containing polycarboxylic acids such as nitrilotriacetic acid, which is a reagent used for the decontamination of nuclear reactor coolant systems. Hence, three commercial corrosion inhibitors (Philmplus 5K655, Prosel PC-2116, and Ferroquest LP7203) were evaluated for high-temperature applications. Preliminary screening of the inhibitors was done by electrochemical techniques, namely, polarization and impedance spectroscopy. Philmplus showed maximum corrosion inhibition efficiency and hence was used for high-temperature investigations. A concentration of 500 mg/L was found to be optimum. The high-temperature dissolution of corrosion product oxides such as magnetite and nickel ferrite that are relevant to nuclear reactors was also carried out in the presence of Philmplus. During the decontamination process, which involves the dissolution of corrosion product oxides, it is desirable to use an inhibitor that will alleviate the corrosion of the underlying base metal without compromising on the dissolution of the oxides present over it. Investigations were also carried out to evaluate hydrazine as a corrosion inhibitor for high-temperature applications; the results obtained were comparable to those of Philmplus. [ABSTRACT FROM AUTHOR]

To support the functionality of RELAP5-3D for best-estimate reactor simulation code, a variety of utility programs were developed at Idaho National Laboratory. Conversion and upgrades of RELAP5-3D to FORTRAN 95 required the upgrades of these utility programs as well. Pygmalion (Pygi) and the RELAP5-3D Graphical User Interface (RGUI) are two of the utility programs that were upgraded and enhanced. Pygi creates a copy of a RELAP5-3D input file with updated initial condition information. From the restart/plot file of the RELAP5-3D run of an input deck, Pygi obtains the final conditions for each component. It creates a new input file that replaces the original input file values with updated conditions. This provides an accurate and efficient means of creating new input decks with steady-state input conditions. RGUI is an alternative to the command line interface for performing RELAP5-3D-related work. RGUI provides single-interface access to other RELAP5-3D tools such as Pygi. It is highly configurable, allowing the user to customize the environment. The interface runs on both Linux and Windows. This suite of utility programs is continually being enhanced to provide better support for RELAP5-3D users. This technical note provides details of Pygi and RGUI functionality. Future plans for enhancements are also included. [ABSTRACT FROM AUTHOR]

The RELAP5-3D system analysis code is a generic and flexible analysis program that is used for the analysis of steady-state and transient scenarios of nuclear reactors. While RELAP5-3D has been successfully applied to a very large range of problems, other analysis tools exist that provide either additional accuracy or capability. This paper describes an extensible integrated code system that utilizes the RELAP5-3D analysis code in conjunction with the R5EXEC executive. A description of the different types of time-step algorithms and data transfer methods is provided. Discussions of the various strengths and weaknesses of the coupled code system are provided. Finally, examples of how the coupled code system has been exercised are included. [ABSTRACT FROM AUTHOR]

In Switzerland 40% of the electricity generation is produced by nuclear power. With all five reactors being already beyond their 30th year of operation, Nagra (National Cooperative for the Disposal of Radioactive Waste) in collaboration with the utilities periodically contributes to the Swiss Nuclear Power Plant (NPP) decommissioning cost studies. These studies are of relevance to the estimation of the financial input of the utilities to the Swiss decommissioning fund and the planning of decommissioning activities. During reactor operation, a fraction of the neutrons produced in the reactor core will escape the core boundaries and eventually interact with the surrounding matter. The most heavily irradiated components are located in the proximity of the reactor core [e.g., core baffle, core support plates, core barrel, and reactor pressure vessel (RPV)]. Neutrons will also stream in farther ex-RPV areas and activate components such as the reinforced concrete bioshield. Decommissioning costs are dependent, inter alia, on the radioactive waste volumes and on the corresponding isotopic inventories. Neutron-activated components are the main source of radioactivity within a NPP under immediate dismantling (i.e., spent fuel has been removed from the reactor). Reliable neutron transport and activation calculations are, therefore, essential for the estimation of radioactive waste volumes, the selection of an optimal dismantling strategy, the development of the radioactive waste packaging and logistics concept, and consequently for the estimation of the decommissioning costs. In this context, Nagra has developed a state-of-the-art NPP activation calculation sequence that enables the radiological characterization of the Swiss NPPs. This paper focuses on aspects relevant to the neutron transport calculations for a Swiss pressurized water reactor. More specifically, the MCNP5 modeling approach together with the use of the ADVANTG hybrid, variance-reduction acceleration code, is outlined. Furthermore, the validation of the neutron transport calculations with an in situ full-cycle foil activation campaign is presented. [ABSTRACT FROM AUTHOR]

Fuel used in nuclear reactors contains fissile material. The fission process releases a huge amount of energy, and hence, the fissioning components must he held in a robust form capable of enduring high operating temperatures and an intense radiation environment. The shape and integrity of the fuel structures must be maintained over a period of several years within the reactor core to prevent the leakage of fission products into the reactor coolant. Further, the fuel rods must be in a nondistorted state for proper alignment in the fuel assembly to ensure proper fuel bundle power distribution. Improper core power distribution can breach the safety and operational limits on fuel and channel powers. The strategy discussed includes the methodology to verify the fuel assembly using image processing techniques. The methodology uses the Radon transform and contains four phases: image reading, preprocessing, Radon transform, and verification. The approach has been validated on 1026 fuel assemblies of a nuclear power plant, for which experimental results are shown. [ABSTRACT FROM AUTHOR]

Integral pressurized water reactors are innovative reactors in which all of the components typically associated with the nuclear steam supply system of a nuclear power station are located within the reactor pressure vessel. In order to facilitate this modification in large [∼1000-MW(electric)] light water reactors (LWRs), compact heat exchangers such as microchannel heatexchangers must be used. Previous attempts at using microchannel heat exchangers were unsuccessful since they are prone to vapor locking and crud blockage when the primary coolant boils. Therefore, the authors propose the use of a flashing drum to facilitate boiling in conjunction with a primary microchannel heat exchanger for a large integral LWR. The integral inherently safe light water reactor (I2S-LWR) is used as a basis for the implementation of this novel concept. The high-temperature, high-pressure secondary water generated in the secondary loop through heating in the microchannel primary heat exchanger of the I2S-LWR is sent to a flashing drum where 99.9% pure vapor is extracted and sent to the turbines. This prevents boiling in the primary heat exchanger that in turn reduces crud deposition, flow instabilities, and the potential for channel blockage or vapor locking in the small channel sizes of microchannel heat exchangers. The benefits and disadvantages of this approach are presented in this paper. Unfortunately, this innovative approach to nuclear steam generation for integral LWRs is challenged by a potential decrease in thermodynamic efficiency. Therefore, a sensitivity study is presented that explores the impact of several design variables on the thermodynamic efficiency of the plant. As part of this study, a simple and, a complex Rankine cycle were modeled in order to determine the impact that system design modifications can play in recovering thermodynamic efficiency lost by the steam drum. Both cycles utilize turbines, condensers, and condensate/recirculation pumps, while the complex Rankine cycle utilizes a four-stage turbine with subsequent separation and open feedwater heaters. The optimized efficiencies for the simple and complex Rankine cycles are 31% and 33%, respectively, indicating that additional system enhancements to the power conversion system could compensate for the inclusion of a flashing drum. [ABSTRACT FROM AUTHOR]

Simulation results are presented for a design strategy that seeks to achieve inherent control and passive safety for liquid-metal advanced small modular reactors. The approach places an increased reliance on passive feedbacks to regulate plant operation. A reference liquid-metal reactor design is defined to serve as a baseline against which innovative design concepts can be compared with respect to operational performance. The definition assigns values to key plant parameters related to materials type, component data, system configuration (loop versus pool type), fuel cycle (burner versus breakeven versus breeder), and balance of plant. The reference design represents the state of the art of conventional fast reactor technology in terms of economics of electricity production, use of active control systems, and standard operation (e.g., refueling every 2 to 3 years). Innovative design features and associated control strategies are then investigated for reducing the size of upset imitators and for improving also the safety of the inherent response to the initiator. Initiators include failures of active systems and operator errors. At the same time the ability of the modified plant to meet normal grid demands subject to constraints on temperature rates of change is assessed. Results presented indicate that operational performance can be maintained while active system initiator size is reduced resulting in improved safety. Essentially, the innovations introduce inherent feedback mechanisms that serve to reduce the magnitude of the control action of the active control systems. [ABSTRACT FROM AUTHOR]

In the case of a severe accident in a core resulting from unprotected loss of flow (ULOF) or unprotected transient overpower, damage can propagate from subassembly to subassembly and produce a whole-core-scale molten pool. Because the core is not in its most reactive configuration, a massive collapse of the molten material could result in a rapid supercritical condition with release of a large amount of energy. However, timely and sufficient fuel relocation outside the core by dedicated means could prevent any risk of recriticality and accident escalation. Based on a reference 1500-MW(electric) sodium-cooled fast reactor design, this paper describes the main results obtained in evaluating the recriticality potential of the European Sodium Fast Reactor (ESFR) core and conditions for its elimination during a ULOF-type transient. This study has been carried out in the fi-ame of the Collaborative Project on European Sodium Fast Reactor of the 7th Framework Programme Euratom. The numerical analyses carried out in the present work allow one to estimate the amount of fuel mass that has to be removed from the core in order to maintain it in subcritical conditions, preventing the formation of a critical pooh Requirements for successful application of this approach, in terms of the negative reactivity insertion rate by fuel relocation and timing of discharge from the core, are derived. [ABSTRACT FROM AUTHOR]

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development!Nuclear Energy Agency (OECDINEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters-Kcff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution-were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly-level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.O data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared. [ABSTRACT FROM AUTHOR]

We present a stability test of the explicit Euler and predictor-corrector-based coupling schemes in Monte Carlo burnup calculations of the gas fast reactor fuel assembly. Previous studies have identified numerical instabilities of these coupling schemes in Monte Carlo burnup calculations of thermal-spectrum reactors due to spatial feedback-induced neutron flux and nuclide density oscillations, where only sufficiently small time steps could guarantee acceptable precision. New results suggest that these instabilities are insignificant in fast-spectrum assembly burnup calculations, and the considered coupling schemes can therefore perform well in fast-spectrum reactor burnup calculations even with relatively large time steps. [ABSTRACT FROM AUTHOR]

Pressure waves might he expected in a nuclear reactor system due to a sudden rupture of a pipe or to quick opening or closure of a system valve. Once generated, they can result in large mechanical loads on the reactor pressure vessel internal structures and pipelines, threatening their integrity. This kind of phenomenon is an important issue and a limiting accident case for nuclear power plant safety, which requires an extensive analysis to ensure plant safety. To study these phenomena, four pressure wave propagation (PWP) tests have been performed in the PMK-2 test facility in MTA-EK. In addition, the first one of the four tests has been used to assess the capability of the MARS-KS code in simulating PWP phenomena. Then, an input model representing the PMK-2 test facility was developed to simulate the tests. Herein, the MARS-KS code simulation results are compared with the test results for the first PWP test. The comparison shows that the MARS-KS code can simulate the PWP frequencies and pressure wave peaks well. After this qualified assessment, the MARS-KS code is then deployed to conduct a sensitivity analysis on the effect of the break size, break opening times, initial coolant conditions, and existence of the pressurizer on the PWP phenomena. The sensitivity analysis on the break opening times shows that the pressure wave amplitude is relevant to the break opening times and that the shorter the break opening time is, the faster the pressure depressurizes. The sensitivity analysis on the effect of the break sizes shows that the larger the break size is, the higher the pressure peak is. And, there is little effect of initial coolant pressure and temperatures and isolation of the pressurizer. [ABSTRACT FROM AUTHOR]

The recent interest in the Small Modular Reactor (SMR) for its potential increased economic competitiveness has focused attention in part on reducing operational costs to offset those plant costs that do not benefit from the economies of scale of large traditional units. Plant operation and maintenance economics are significantly driven by plant availability, which can be enhanced by means of innovative control strategies by avoiding unnecessary plant or unit trips. In this context, an effective strategy for achieving fast runback of a sodium-cooled SMR has been developed. In this work, after having defined and modeled a suitable control strategy by adopting the Petri nets formalism, a Model-based Predictive Control regulator has been developed in order to reduce as promptly as possible the power level, without scramming the reactor (fast runback) and possibly limiting the control rod contribution. Such flexibility could lead to significant savings in the operational costs of the reactor while also improving the system availability. The proposed procedure has been characterized by simulating the operational transients on both an oxide-fueled reactor and on a metal-fueled reactor, comparing the responses of the two different configurations and the respectively needed control rod contribution. [ABSTRACT FROM AUTHOR]

In the advanced passive pressurized water reactor, a passive containment cooling system (PCCS) has been adopted to cool the containment--comprising a cylindrical steel vessel--during postulated accidents, whereby the decay heat is removed through water film evaporating enhanced by air cooling outside the containment. In this study, an integrated safety analytical code is used to study the heat removal capacity of PCCS during severe accidents and its influence on severe accident management measures. The coupled analytical model includes the reactor cooling system, engineered safety features, containment system, and PCCS. Containment responses during typical design-basis accidents and integrated severe accident scenarios are calculated and validated using a design control document and probabilistic risk assessment, respectively. Four typical severe accident sequences that contribute to core damage frequency or containment high pressure are selected to evaluate the containment response. The results show that the containment pressure can be controlled at a relatively low level within 72 h with the heat removal by PCCS. Analysis of the effects of PCCS water cooling recovery during the late period of the accident sequence in severe accident management guidelines alerts as to the risk of hydrogen combustion after breaking the steam-inert atmosphere inside containment. Moreover, sensitivity analysis has been performed to study the influence of the water film coverage rate and environmental air temperature, and it shows that a decrease of the water film coverage rate and an increase of the environmental air temperature reduce the PCCS cooling capacity. [ABSTRACT FROM AUTHOR]

New materials are being considered for fuel, cladding, and structures in next-generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high-temperature irradiation. To accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide these data, programs such as the Advanced Test Reactor (ATR) National Scientific Users Facility (NSUF) have funded researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) to develop several instrumented test rigs to obtain data in real time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in ATR. This technical note reports the status of INL efforts to develop and evaluate prototype test rigs that rely on linear variable differential transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower-flux materials testing reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in the length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-cladding interaction, and crud buildup. [ABSTRACT FROM AUTHOR]

A new daily as-run thermal analysis was performed at the Idaho National Laboratory for the advanced gas cooled reactor (AGR) test experiment number two (AGR-2) in the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment due to graphite shrinkage resulting from neutron damage. The purpose of this analysis was to calculate the daily average temperatures of each TRISO (tristructural isotropic)-particle fuel compact. A steady-state thermal analysis was performed daily for each capsule with the commercial finite element heat transfer code ABAQUS. These new thermal predictions show the compact fuel temperature dependence on the variable gas gap method. Comparison between measured and calculated temperatures is discussed. [ABSTRACT FROM AUTHOR]

Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and postirradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory and then later shipped to the Hot Fuel Examination Facility located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and nonfueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads. [ABSTRACT FROM AUTHOR]

The release behavior of radioactive materials from high active liquid waste (HALW) has been investigated under boiling accident conditions. Results of the experiment using a nonradioactive simulated HALW found Ru to be a volatile element under the accident conditions and to be released into the gas phase in the form of both mist and gas. The Ru release rate and the apparent Ru volatilization rate constant were obtained under the boiling conditions of simulated HALW. The other fission product elements such as Cs were found to be nonvolatile and to be released into the gas phase in the form of mist. The mist size distribution near the surface of the simulated HALW in the reactor vessel was found to range from 0.05 to 20 pm with a peak diameter of ~2 pm. [ABSTRACT FROM AUTHOR]

Nuclear reactor systems codes have the ability to model the system response in an accident scenario based on known initial conditions (ICs) at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermomechanical fuel rod response models needed for best-estimate prediction of fuel rod failure. Alternatively, the reverse can be said about fuel performance codes; they can lack the ability to capture and model the thermal-hydraulic (T-H) influence of adjacent fuel rods and the rod's location in the reactor core. This work analyzes the limitations in using fuel performance codes to represent in-reactor conditions as determined by full-core T-H codes. The codes used in this analysis are the U.S. Nuclear Regulatory Commission s steady-state fuel performance code FRAPCON-3.5 and T-H code TRACE-V5P3. In order to assess the impact of the limitations found in the codes, several modifications were made to all of the codes to improve code-to-code consistency. The modifications to the fuel performance code include adding the ability to model gamma-ray heating and providing realistic core coolant conditions. The T-H code modifications include adding the ability to model the fuel with axially varying burnup-dependent fuel and cladding dimensional changes and corrosion characteristics. The fuel in a Westinghouse four-loop pressurized water reactor was modeled to assess the impacts these modifications have on fuel performance and ICs for transient analysis. The results of this study show that current modeling assumptions (and limitations) can yield both conservative and nonconservative results on several important licensing criteria. [ABSTRACT FROM AUTHOR]

Source term is an important issue in safety assessment of nuclear power plants. Therefore, modeling of particulate concentration in reactor coolant systems during normal operation and hypothesized accidents is of continuing interest. We report here on exploration of a numerical solution of the Reeks-Hall equation with the use of the fractional resuspension rate in its original integral form. The numerical results for particulate concentration are compared with those obtained from the exact expression given by Williams and experimental data provided by Wells et al. The numerical results agree very well with exact results and also agree well with the data of Wells et al. Applications of the numerical method to problems with a time-dependent resuspension rate (for which exact solutions are not available) are explored, and some typical results are reported. The numerical method will be useful for verifying approximate techniques that are used for aerosol modeling in nuclear source term computer programs. [ABSTRACT FROM AUTHOR]

Empirical models are applicable over limited ranges of their predictor variables. The space defined by those ranges, the application domain, is the entire space over which the empirical model is applied. One important assumption is that the model's predictive behavior is consistent over the entire application domain. This assumption is commonly made for critical heat flux (CHF) models when they are applied in reactor safety analysis. The intention of this work is to demonstrate that the current assessment methods used to justify this assumption may not always identify subregions in the application domain where the model's predictive capability is degraded. This is accomplished by intentionally placing a nonconservative subregion in a CHF model and demonstrating that the current assessment methods are unable to identify that nonconservative subregion. As the existence of a nonconserwative subregion may impact reactor safety analysis, a new method is proposed that does identify the nonconservative subregion. This new method is a multidimensional approach capable of demonstrating if the CHF model's predictive behavior is likely due to random effects or is due to degraded predictive capability. [ABSTRACT FROM AUTHOR]

Risks related to sloshing of liquid metal coolant due to seismic excitation need to he investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided. [ABSTRACT FROM AUTHOR]

The Fukushima Daiichi nuclear accident showed that severe events are real-not virtual-threats. In both pressurized water reactor and boiling water reactor nuclear plants, the containment is the last resort against radiation leakage, with the integrity of containment being directly linked to nuclear safety. To maintain the containment below Factored Load Category for 24 h and beyond after core damage, external cooling measures such as a gas vent system and an emergency containment spray backup system have been mandated. With these measures, it is difficult to completely satisfy all the requirements for proper cooling and minimized penetration and leakage while maximizing reliability and meeting the need for a passive design. Accomplishing all of these requires substantial repair and maintenance costs; thus, many difficulties can be anticipated with their implementation. This study proposes the installation of several multipod heat pipe assemblies; such an assembly includes an adiabatic region consisting of one large cylindrical structure penetrating the containment dome, a boiling region, and a condenser region consisting of many pipes that serve as the ultimate heat sink to discharge the decay heat energy from the containment with no radiation leak. Such installation will dramatically improve the nuclear safety in the event of a severe accident. [ABSTRACT FROM AUTHOR]

The supercritical carbon dioxide (S-CO2) recompres-sion cycle is a power conversion cycle compatible with intermediate-temperature nuclear reactors. The main advantage of the S-CO2 cycle is relatively high efficiency (~47% at the turbine inlet temperature of 650°C). The dynamic characteristics and control of this cycle remain areas of active research because of the cycle's unique features, in particular, large fluid property changes near the critical point. This paper reports the conceptual development of a dynamic S-CO2 recompression cycle controller designed to efficiently respond to a demand for reduction of generator electric power. The S-CO2 cycle generator electric power production can be controlled using either turbine bypass (TB) or mass inventory (MI) controllers. Turbine bypass is a fast response controller, which reduces generator power by opening the TB valve. Mass inventory is a slow response controller, with a time constant an order of magnitude larger than that of the TB controller. The Ml controller reduces generator electric power through decreasing the inventory of CO2 gas in the cycle by pumping some of the gas into a storage reservoir. Both TB and MI controllers operate in conjunction with a precooler temperature controller, which maintains compressor inlet conditions near the critical point. Although using a TB controller allows for quick reduction of the generator electric power, S-CO2 cycle thermal efficiency is reduced during the steady-state operation. Cycle efficiency can be improved if cycle control is transitioned from the TB to the MI controller. However, directly switching from the TB to the MI controller would result in a spike in generator power because of the large discrepancy between the time constants of the two cycle control modes. To address this deficiency, we have designed a mixed-mode (MM) controller to transfer cycle control to MI mode after steady state has been reached in TB mode. In the MM controller, both TB and MI controllers operate simultaneously, thus maintaining nearly constant generator electric power during S-CO2 cycle control transitioning. Design of an MM controller for the S-CO2 cycle does not appear to have been previously reported in literature. To test our controller design, we have petformed proof-of-concept numerical experiments. All controllers in this study were implemented as proportional-integral controllers using the System Control Module (SCM) language. Gain coefficients for all controllers were determined via numerical experiments, in which response of the S-CO2 cycle was calculated with the GPASS (General Plant Analyzer and System Simulator) software package. Gain coefficients and cycle timescales were calculated under idealized conditions of instantaneous measurement response. [ABSTRACT FROM AUTHOR]

A feedforward artificial neural network (ANN) is constructed using select nuclear criticality excursion experiment data sets from the French Consequences Radiologiques d'un Accident de Criticité (CRAC) and SILENE reactor campaigns. The ability to represent initial spike characteristics by an ANN provides a new method that is aligned to excursion data more directly and to a wider variable data set than traditional analytic approaches. The ANN is configured, trained, validated, and tested to 85 unique highly enriched uranium (HEU) excursion experiments, considering six input variables and two output variables (specific power and energy). The fidelity of the ANN is enhanced by normalizing the input and output data. The trained ANN is then used to determine output values for 19 select Kinetic Energy Water Boiler experiments and 14 additional CRAC excursions not used in the ANN construction. Furthermore, the same trained ANN is also used for an extensive comparison (80 cases) for a combination of uranium concentrations, ramp feed reactivity insertion rates, system volumes, and vertical container sizes. The specific spike energy and power ranges determined are bracketed by published experiment results and are more realistically represented than results derived from well-known analytical methods. The ability to predict initial peak fissions by an ANN does not require determining, a priori, a volume-dependent energy quench parameter ("b") specific to HEU solutions. The results derived from the ANN can aid in designing realistic emergency planning constructs or criticality accident alarm system hardware placements without undue penalty for fission source term uncertainties. Neither excursion characteristics after the initial spike nor explicit time dependencies are modeled by an ANN at this time. The extension of the methods presented is left for further work. [ABSTRACT FROM AUTHOR]

The Advanced Test Reactor (ATR) is conducting scoping studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low-enriched uranium (LEU) composition, through the Reduced Enrichment for Research and Test Reactors Program, within the Global Threat Reduction Initiative. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary reactor physics scoping and feasibility analysis of TRIGA fuel within the current ATR fuel element envelope and compares it to the functional requirements delineated by the Naval Reactors Program, which includes >4.8 x 1014 fissions⋅s-1 of 235U in test positions, a fast-to-thermal neutron flux ratio that has a <5% deviation from its current value, a desired steady cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other design parameters outside those put forth by the Naval Reactors Program that are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time. The result of this study demonstrates potential promise for implementation of TRIGA fuel in the ATR from a reactor physics perspective; discussion of observations and limitations are provided herein. [ABSTRACT FROM AUTHOR]

Failed fuel detection for the Japan Sodium-cooled Fast Reactor (JSFR) has been studied. The present JSFR design adopts a selector-valve (SV) failed fuel detection and location (FFDL) system. In this study, a tagging-gas (Tag) FFDL (Tag-FFDL) system has been investigated as an alternative. Although the identification performance of the Tag-FFDL system has been demonstrated in small and medium-sized reactors, the Tag-FFDL system has not been demonstrated yet in a large reactor like JSFR, which has 1500-MW(electric) power and 562 core fuel sub-assemblies. Major issues of the JSFR Tag-FFDL system are affected by high-burnup fuel and large cover gas volume. High-burnup fuel leads to a large change of the isotope ratio, which is important for the detection performance of the Tag-FFDL system. Since the cover gas volume in JSFR is larger than that in previous reactors, the tagging-gas concentration in the cover gas is lower than that in previous reactors. Thus, a requirement of the background value is more strict in JSFR. This study investigates whether two issues of the Tag-FFDL system for JSFR would be solvable. Tag gas isotope change in a high-burnup condition has been evaluated regarding transmutation and fission gas release. Taking into account tag gas isotope change due to the high-burnup conditions and large cover gas space, JSFR tagging gas has been designed. The investigation results showed that the JSFR FFDL system can provide an identification capability for 672 subassem-blies, which is larger than the number of subassemblies in JSFR combining tagging-gas and burnup estimation. In addition, an allowable background concentration of natural Kr and Xe in the cover gas has been evaluated. [ABSTRACT FROM AUTHOR]

This technical note introduces a numerical procedure that is efficient for calculating the solution for the fractional order nonlinear neutron point-kinetics equation in nuclear reactor dynamics. The explicit finite difference method (EFDM) is applied to solve the fractional order nonlinear neutron point-kinetics equation with Newtonian temperature feedback reactivity. This nonlinear neutron point-kinetics model has been analyzed in the presence of temperature feedback reactivity. The numerical solution obtained by EFDM is an approximate solution that is based on neutron density, precursor concentrations of multigroup delayed neutrons, and the reactivity function. The method is investigated using experimental data, with given initial conditions along with Newtonian temperature feedback reactivity. From the computational results, it can be shown that this numerical approximation method is straightforward and effective for solving fractional order nonlinear neutron point-kinetics equations. Numerical results citing the behavior of neutron density for different types of fractional order are presented graphically. [ABSTRACT FROM AUTHOR]

A program is underway to convert the current high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel in the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology. A RELAP5 model has been developed to analyze postulated accidents in the NBSR with the present HEU fuel and a proposed LEU fuel. The model includes the reactor vessel, primary pumps, shutdown pumps, various valves, heat exchangers, and average and hottest fuel elements and flow channels in the region where flow enters through an inner plenum (6 fuel elements) and a region where flow enters through an outer plenum (24 elements). The equilibrium cycle power distributions in the fuel elements were determined based on three-dimensional Monte Carlo neutron transport calculations performed with the MCNPX code. In this paper we discuss safety analyses conducted for the loss-of-flow accidents resulting from either loss of electrical power or inadvertent throttling of flow control valves at the inlets to the inner and outer plena. The analysis shows that the fuel conversion will not lead to significant changes in the safety analysis and that the calculated maximum clad temperatures, minimum critical heat flux ratios, and minimum onset of flow instability ratios assure that there is adequate margin to fuel failure. [ABSTRACT FROM AUTHOR]

In response to the accident at the Fukushima Daiichi (IF) nuclear power plant, designers of the 2010 version of the Japan sodium-cooled fast reactor (JSFR) have been studying the robustness of JSFR's fuel handling system (FHS) against an earthquake and a tsunami. In the earthquake evaluation, the margin of seismic resistance and sloshing in the ex-vessel fuel storage tank (EVST) against an earthquake of the IF-envelop condition were estimated. In the tsunami evaluation, for the case of loss of emergency power supply, a scenario is studied where fuel subassemblies are led to a stable cooling state, and potentialities for the cooling system are examined. As a result, it is shown that the original design of the JSFR FHS already had the potential to prevent the release of radioactive material. Additionally, some design improvements of the EVST cooling system are introduced. [ABSTRACT FROM AUTHOR]

As a result of the mass balance analysis for fission products (FPs) in the contaminated water accumulated in the Fukushima Daiichi nuclear power plant in the aftermath of the accident, it was concluded that the short-term cesium source was 50% of the total amount of cesium accumulated in the reactor, which was estimated with the SAMPSON severe accident analysis code, while the short-term source of tritium was 26% of the total tritium that was released during the U02fuel meltdown. The FP concentrations in the contaminated water during the 2 yr since the accident were determined by short-term FP sources, while their saturated concentrations, due to a balance between the release from the reactor and the cleanup, were determined by long-term FP sources. A multinuclei removal system, ALPS, has been operated at the plant to clean up the contaminated water to reduce FP concentrations to a level that is permitted for release to the environment. Tritium concentration in the contaminated water decreased a little during the 2 yr since the accident primarily from dilution by mixing groundwater rather than intentional removal. Industrial-scale removal of tritium from the huge amount of contaminated water is almost impossible using isotopic separation methods. Storage of tritiated water for a long period presents a high risk for leakage to the environment. As the most realistic procedure, the authors propose that the tritium be diluted with plenty of seawater to the natural background level and then be released into the ocean with continuous monitoring at the release point. [ABSTRACT FROM AUTHOR]

Molten FLiNaK salt [46.5%LiF-11,5%NaF-42%KF (mol%)] has been proposed for use as a secondary reactor coolant and medium for transfer of high-temperature process heat from nuclear reactors to chemical plants. Two alloys--Hastelloy-N superalloy (Hastelloy-N) and Type 316L stainless steel alloy (316L steel)--were exposed to molten FLiNaK salt in a 316L steel crucible under argon cover gas at 850°C for 1000 h. Graphite was also introduced into the test with the goal of studying the corrosion behavior of relevant reactor material combinations. The results show that corrosion of 316L steel occurred primarily through surface depletion of Cr. Contrarily, Hastelloy-N experienced weight gain due to the electrochemical plating of corrosion products, Fe and Cr, derived from the 316L steel crucible. The graphite sample enhanced the corrosion of the 316L steel sample and crucible, which induced the formation of (Cr,Fe)7C3 and (Mo,Cr,Fe)2C carbides on the surface of graphite. These carbide formations were attributed to the nonelectric transfer between 316L steel and graphite. Besides reducing the availability of chromium to plate, the presence of graphite did not change the basic corrosion of the 316L steel and plating process of Hastelloy-N. [ABSTRACT FROM AUTHOR]

This paper presents a preliminary systems analysis related to most of the currently proposed enhanced accident-tolerant fuel and cladding concepts: fully ceramic microencapsulated fuels, uranium-molybdenum fuels, uranium-nitride fuels, uranium silicide fuels, silicon carbide cladding, advanced steel cladding, and molybdenum cladding. The benefits drawn from the implementation of demonstrated accident-tolerant fuels on the future development of nuclear energy generation as well as public acceptance are difficult to quantify but would probably be very significant. The potential impacts of these innovative light water reactor fuels on the front end of the fuel cycle, on the reactor operation, and on the back end of the fuel cycle are succinctly described. [ABSTRACT FROM AUTHOR]

Recent advances in technologies for harvesting waste thermal energy from ambient environments present an opportunity to implement truly wireless sensor nodes in nuclear power plants. These sensors could continue to operate during extended station blackouts and during periods when operation of the plant's internal power distribution system has been disrupted. The energy required to power the wireless sensors must be generated using energy harvesting techniques from locally available energy sources, and the energy consumption within the sensor circuitry must therefore be low to minimize power and hence the size requirements of the energy harvester. Harvesting electrical energy from thermal energy sources can be achieved using pyroelectric or thermoelectric conversion techniques. Recent modeling and experimental studies have shown that pyroelectric techniques can be cost-competitive with thermoelectrics in self-powered wireless sensor applications and, using new temperature cycling techniques, have the potential to be several times as efficient as thermoelectrics under comparable operating conditions. The development of a new thermal energy harvester concept, based on temperature cycled pyroelectric thermal-to-electrical energy conversion, is outlined. This paper outlines the modeling of cantilever and pyroelectric structures and single-element devices that demonstrate the potential of this technology for the development of high-efficiency thermal-to-electrical energy conversion devices. [ABSTRACT FROM AUTHOR]

The purpose of this study was to design a robust test facility for a small space nuclear power system and model its physical behavior under different scenarios. The test facility will be used to simulate a 1- to 10-kW(electric) nuclear reactor, its electrical generation, and heat removal capabilities. This simulator will be used to explore, test, and understand the steady-state and transient operation capabilities of small space nuclear power systems. Currently, the system is planned to operate on a variable, electrical heat source directly connected to heat pipes. The heat pipes are to be stainless steel with a water working fluid. These heat pipes will then be connected to a power conversion simulator or actual power conversion technologies. The power conversion simulator is connected to a radiator using a water-based heat pipe network using fins and connecting plates in a cylindrical geometry. Modeling of the facility was performed using the SolidWorks Flow Simulation package. Flow Simulation was used to analyze startup, heat pipe failures, and loss of power conversion with the end goal of finding safe operational transient scenarios for the transient test facility. [ABSTRACT FROM AUTHOR]

The subcritical advanced burner reactor (SABR) concept, which combines IFR-PRISM fast reactor technology and the ITER tokamak fusion physics and technology in a burner reactor for the transmutation of transuranics, has been adapted for a subcritical advanced breeder reactor (SABrR) that produces plutonium. It is found that basically the same fission and fusion technology, geometry, and major parameters as used in SABR can be used to achieve a significant fissile production rate (fissile breeding ratio ≈ 1.3) while maintaining tritium self-sufficiency (tritium breeding ratio >1.15) [ABSTRACT FROM AUTHOR]

Because of the Great East Japan Earthquake, and the resulting tsunami, which occurred on March 11, 2011, a serious accident occurred in Units 1, 2, and 3 of the Fukushima Daiichi nuclear power station. Since the accidents, data from interviews with operators and on-site surveys have been continuously compiled. Based on the data, a plant-state analysis has been conducted using the severe accident analysis code MAAP (Modular Accident Analysis Program). Parallel to the MAAP analysis, the responses of the plant to site operations, such as water injection, are analyzed, and core conditions are comprehensively evaluated. According to the evaluation, in Unit 1, it is presumed that almost no fuel was left at the original position; it was molten and moved downward. The fuel likely damaged the reactor pressure vessel (RPV), and it is assumed that most of it had dropped to the primary containment vessel (PCV) pedestal. In Units 2 and 3, it is presumed that some of the fuel was left at the original position and the rest dropped to the bottom of the RPV or to the PCV pedestal. In the MAAP analysis, the behavior of the plants before core melt is reproduced. However, RPV damage of Units 2 and 3 does not occur in the MAAP analysis, which is contrary to the observed facts. This shows that the analysis capability of the current MAAP code is limited. Therefore, by developing severe accident analysis codes to achieve higher levels of accuracy and by evaluating the plant responses to site operation, we will continue to obtain a clear picture of the states inside the reactor so that fuel debris can be removed. [ABSTRACT FROM AUTHOR]

The risk-dominant containment failure modes of a pressurized water reactor are reassessed using the current state of knowledge for the phenomena that contribute to these failure modes. Our review concludes that some mechanisms that were considered as having the potential to result in containment failure at the time of NUREG-1150, such as in-vessel steam explosions and vessel launch (i.e., the alpha-mode containment failure), have subsequently undergone sufficient review and can be excluded from further consideration. For other phenomena, such as high-pressure melt ejection (HPME) and combustible gas explosions, our review concludes that substantial uncertainties still exist with regard to modeling in system level codes; for combustion events, careful consideration is still required when making severe accident management decisions. With regard to HPME, sensitivity studies have been performed with the MELCOR computer code to address the effects of modeling uncertainties on containment loading. Sensitivity studies using MELCOR have also been performed with regard to combustion events to examine gas generation, the effect of containment cooling on the potential for deflagrations, and the combustion load on containment. Combustion loads are compared to the NUREG-1150 containment fragility curve to assess the likelihood of containment failure. Our MELCOR analyses agree with the NUREG-1150 assumption that insufficient hydrogen is generated in-vessel to result in containment failure. Sensitivity studies regarding the rate and timing of reflooding a degraded core do not indicate a significant effect on hydrogen production in-vessel or a significant challenge to containment integrity regarding HPME. However, it is observed that recovery actions resulting in cooling of the containment atmosphere could result in deinerting the containment and lead to a sufficiently energetic combustion event that can fail the containment. [ABSTRACT FROM AUTHOR]

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty due to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. The MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012. [ABSTRACT FROM AUTHOR]

Methods to remedy the ill-posedness of the basic one-dimensional two-fluid model, which is widely used in nuclear reactor safety codes, have been the subject of considerable study. Both of the two prevalent methods have drawbacks. Unconditional hyperbolization uses nonphysical constitutive relations to create a well-posed two-fluid model that is hyperbolic over all flow conditions. However, when the model is hyperbolized, it is also stabilized, which is not a universal property of two-phase flows. The second method, the preferred method of the U.S. Nuclear Regulatory Commission safety codes, is to simply use a first-order upwind numerical method that relies on numerical viscosity to regularize the ill-posedness of the model by damping the short-wavelength instabilities. Unfortunately, the scale of the "short wavelength" is related to a particular numerical grid or discretization. Because of the consistency of the numerical method, in the limit of an infinitely resolved grid, i.e., Δx, Δt → O, the numerical viscosity vanishes, as does its regularization effect. This results in a somewhat heuristic user guideline that suggests a lower limit on the grid size based on a cross-sectional dimension Δx ≥ DH that is a combination of the long-wavelength assumption and experience. However, a cutoff wavelength achieved by numerical viscosity is not set by the grid size alone but also depends on the time step, the material, and the flow properties, as demonstrated with avon Neumann stability analysis. This can create poor resolution in areas where numerical stability may not be a substantial problem, unless the guideline is intentionally violated. Additionally, strict observance of this limit makes verification by convergence difficult or impossible. Therefore, it is proposed that an artificial viscosity be prescribed explicitly, i.e., independently of any particular numerical method or grid. An artificial viscosity model is derived that prescribes exactly a cutoff in the linear stability growth rate at a specified wavelength, e.g., λ0 = 2DH, consistent with the aforementioned user guideline. It is shown, using the water faucet problem, that the proposed artificial viscosity model can be used to remove the high-frequency component of the solution without limiting the resolution of the grid. Furthermore, the solution also converges, which was not the case without the artificial viscosity. [ABSTRACT FROM AUTHOR]

A dual measurement system for monitoring the simultaneous positions of multiple radioactive tracer pebbles in scaled pebble bed reactors (PBRs) has been developed and benchmarked to the prototype stage. The first system (the collimated system) is an updated version of a previously developed system that is now a completely automatic system that uses three collimated directionally variable Nil detectors that are programed to continuously search for a maximum counting rate from a single radioactive pebble. This system can be used by itself when a single radioactive tracer pebble is of interest and the pebble is relatively slow moving. In the present case, its primary use is to provide an independent measurement of the position of a stationary tracer pebble that is used to provide a point for calibration of the second system. The second system (the uncollimated system) is a modified version of a multiple uncollimated NaI detector system commonly called CARPT. The modified version involves those changes necessary to allow for use of the entire gamma-ray spectra for the inverse problem instead of only the gamma-ray full energy peaks. This allows one to use multiple radioisotopes each in a different tracer pebble so that up to ten individual tracer pebbles can be followed simultaneously with the best possible accuracy. The inverse problem is treated with the Monte Carlo library least-squares approach in which Monte Carlo-generated library spectra for each radioisotope are made available for a complete range of reference positions within the scaled PBR. Then, any unknown total gamma-ray spectra can be analyzed in an iterative fashion with the radioisotope library spectra to yield the position of all the radioisotope tracer pebbles. The scaled PBR used was a 30-cm-high and 30-cm-diam circular cylindrical section on the top and a cone with a 25-deg angle on the bottom. The pebbles are 1.2-cm glass marbles. Results have been obtained with both single tracer radioisotope marbles and multiple tracer radioisotope marbles, simultaneously. [ABSTRACT FROM AUTHOR]

Reactivity insertion accidents have been analyzed for the 20-MW D20-moderated research reactor (NBSR) at the National Institute of Standards and Technology (N1ST). The analysis has been carried out for the present core, which contains highly enriched uranium fuel, and for a proposed equilibrium core with low-enriched uranium fuel. The time-dependent analysis of the primary system is performed with a RELAP5 model that includes the reactor vessel, primary coolant pump, heat exchanger, fuel element geometry, and flow channels for both the 6 inner and 24 outer fuel elements. Postprocessing of the simulation results has been conducted to evaluate minimum critical heat flux (CHF) ratio and minimum onset of flow instability (OFI) ratio using the Sudo-Kaminaga correlations and Saha-Zuber criteria, respectively. Evaluations are carried out for the control rod withdrawal start-up accident and the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that no damage to the fuel will occur and there is adequate margin to CHF and OFI because of sufficient coolant flow through the fuel channels and the negative reactivity insertion due to scram. [ABSTRACT FROM AUTHOR]

The fleet of research and training reactors is aging, and no new research reactors are planned in the United States; thus, there is a need to expand the capabilities of existing reactors to meet users' needs. To address these needs, the Colorado School of Mines added a neutron beamline facility to the U.S. Geological Survey TRIGA Reactor (GSTR), a 1-MW(thermal) Mark-I TRIGA reactor located at the Denver Federal Center in Lakewood, Colorado. The original GSTR design did not include any beam ports, and future research efforts will benefit from a neutron beam at the GSTR. Adding new beamline facilities to existing research reactors is both rare and challenging, and this paper describes the design and installation of a new neutron beamline facility at a Mark-I TRIGA reactor with no existing beamline facilities. The design and construction of a radiation beamstop for the new beamline is described in detail. A neutronics model of the neutron beam provides researchers with a useful tool for experiment design. The new neutron beam has a measured length-to-diameter ratio of 200 ± 10, a neutron flux of 2.2 x 106 ± 6. 4 x 105 n/cm²-s, and an average cadmium ratio of 7.4 using copper, gold, manganese, and indium foils. [ABSTRACT FROM AUTHOR]

The sulfur-iodine and hybrid-sulfur thermochemical cycles that can utilize high-temperature heat from advanced nuclear reactors have shown promise economically for large-scale production of hydrogen from water. Both of these cycles employ a step to decompose sulfuric acid to sulfur trioxide by heating it above 723 K followed by the catalytic decomposition to sulfur dioxide at a temperature >1073 K depending on the catalyst used. Successful commercial implementation of these technologies is dependent on the development of suitable materials for use in these highly corrosive environments. In this study, a laboratory-scale superheater/decomposer was constructed and used to study the corrosion resistance of natural diamond, synthetic diamond films treated with boron and titanium, silicon carbide, quartz, aluminum nitride, INCONEL, and platinum to sulfuric acid and SO3. However, it appeared that some of these materials catalyzed SO3 to SO2 and O radicals, which also attacked these materials, increasing their corrosion rates. Natural diamonds, synthetic diamond films (treated with boron and titanium), aluminum nitride, and INCONEL have unacceptable corrosion rates above 873 K. Both the boron- and titanium-treated diamond samples completely disintegrated at temperatures >973 K. The high corrosion rates may have resulted from carbons in diamond having a higher preference for oxygen free radicals that were formed during the decomposition process. Oxygen free radical concentrations increased as a function of the increasing temperature. The present study showed that silicon carbide had the best corrosion resistance over the range of conditions at which the superheater would operate. Quartz was also corrosion resistant but became brittle after 30 h of exposure to this harsh environment. Platinum, used as a catalyst to reduce the decomposition temperatures, exhibited almost no corrosion when exposed to decomposition products. However, platinum did corrode when exposed to liquid sulfuric acid at high temperatures. [ABSTRACT FROM AUTHOR]

The validation of computer codes related to the thermal-hydraulic analyses of nuclear reactors is a challenging undertaking because of the complexity of the phenomena involved, e.g., boiling, condensation, and mixing. In the frame of the ongoing validation of the best-estimate system code TRACE, the present paper focuses on the phenomena taking place during the quenching of the hot surface of the fuel rod simulator with cold water. Since TRACE describes the physical phenomena with empirical correlations derived from experiments, it is necessary to ensure that these correlations are valid if applied to similar experiments but different boundary conditions. By means of an uncertainty and sensitivity study, the influence of the empirical models and their associated uncertainties on selected output parameters is quantified and the parameters with the largest sensitivity are evaluated. [ABSTRACT FROM AUTHOR]

Korea Hydro and Nuclear Power has built a new modular type of CANDU spent fuel bundle dry storage facility, MACSTOR KN-400, at the Wolsong reactor site in the Republic of Korea. Four CANDU reactors operate at the Wolsong site, and the MACSTOR KN-400 has the capacity to store up to 24 000 CANDU spent fuel bundles. The International Atomic Energy Agency safeguards regulations demand an effective method for spent-fuel re-verification at the MACSTOR KN-400 facility in the event of any loss of continuity of knowledge. A radiation signal-dependent spent-fuel re-verification design of the MACSTOR KN-400 is scrutinized through mathematical model development and Monte Carlo radiation transport simulations using the state-of-the-art computer code MCNP. Both gamma and neutron transport simulations for various spent fuel bundle diversion scenarios are carried out for the central and corner re-verification tube structures. The CANDU spent fuel bundles with a burnup of 7500 MWd/tonne U (burned at a specific power of 28.39 MW/tonne) and 10 years of cooling time are considered for the radiation source term. Results of the gamma transport simulations incorporating cadmium-zinc-telluride detectors inside the re-verification tube show that spent fuel bundles diverted from the inner locations of the storage basket cannot be detected by observing a gamma radiation signal change. Neutron transport simulations consisting of a 3He detector inside the re-verification tube show that certain spent fuel bundle diversions could be detected. However, inverse MCNP neutron transport simulations show that the possibility of detecting diversion of ~67% of spent fuel bundles stored in the basket region on the opposite side from the collimator of the re-verification tube is small, assuming a neutron detection counting time of 1 h per re-verification tube. It is also observed that the nondetection probability for most of the diversion scenarios considered is large. Nondetection probability here is defined as the probability of not detecting the diversion of spent fuel bundles from the baskets by observing radiation signal reduction from the removal of the bundles. Containment and surveillance methods are being employed for safeguards purposes at the facility, supplemented by periodic axial profile fingerprinting. However, since the nondetection probability is large for most scenarios, the facility should consider alternatives to this method in case loss of continuity of knowledge occurs. [ABSTRACT FROM AUTHOR]

The development of accelerator-driven systems (ADSs) is motivated by the potential of these machines to reduce the volume and the radiotoxicity of accumulated nuclear waste, more particularly that of minor actinides currently generated by the operation of existing pressurized water reactors. The reduction of both volume and radiotoxicity of nuclear waste is achieved by transmutation and fission of minor actinides into less-active isotopes or shorter-lived by-products. Various technical challenges exist regarding designing reliable and efficient ADSs. The key points are very much linked to the design of the spallation module, the assurance that reactivity remains below criticality under any circumstances, and the accelerator reliability. This paper addresses the latter two challenges im- posed on the accelerator in order to assure safe and reliable ADS operation. It discusses the possibility of performing online absolute reactivity measurements and the limits in the number of allowable accelerator beam trips, which might impede plant integrity and/or plant efficiency. [ABSTRACT FROM AUTHOR]