bluuebunny/crossref_metadata_2025 huggingface dataset:
HTR-TN Achievements and Prospects for Future Developments
Authors: Hittner Dominique , Angulo Carmen , Basini Virginie , Bogusch Edgar , Breuil Eric , Buckthorpe Derek , Chauvet Vincent , Fu¨tterer Michael A. , van Heek Aliki , von Lensa Werner , Verrier Denis , Yvon Pascal
Published: July 2009

It is already 10 years since the (European) HTR Technology Network (HTR-TN) launched a programme for the development of HTR Technology, which expanded through 3 successive Euratom Framework Programmes, with many coordinated projects in line with the strategy of the Network. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR...) that it contributed to safeguard, this programme led to advances in HTR/VHTR technologies and produced significant results, which can benefit to the international HTR community through the Euratom involvement in the Generation IV International Forum (GIF). The main achievements of the European programme performed in complement to national efforts in Europe and already taking into consideration the complementarity with contributions of other GIF partners are presented: they concern the validation of computer codes (reactor physics, system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for the vessel, the direct cycle turbines and the intermediate heat exchanger, graphite...), component development, fuel manufacturing and irradiation behaviour and specific HTR waste management (irradiated fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite to high fluence, fuel material irradiation (PYCASSO experiment), high burn-up irradiated fuel PIE, safety test and isotopic analysis, IHX mock-up thermo-hydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward towards industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission, which promotes a strategy for the deployment of low carbon energy technologies and mentions Generation IV nuclear systems as one of the key contributors to this strategy, HTR-TN proposes to launch a programme for extending the contribution of nuclear energy to industrial process heat applications addressing jointly 1) The development of a flexible HTR able to be coupled to many different process heat and cogeneration applications with very versatile requirements 2) The development of coupling technologies with industrial processes 3) The possible adaptations of process heat applications which might be needed for coupling with a HTR and 4) The integration and optimisation of the whole coupled system. As a preliminary step for this ambitious programme, HTR-TN endeavours presently to create a strategic partnership between nuclear industry and R&D and process heat user industries.
CARBOWASTE: New EURATOM Project on ‘Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste’
Authors: von Lensa Werner , Bradbury D. , Cardinal G. , Eccles H. , Fachinger J. , Grambow B. , Grave M. J. , Marsden B. J. , Pina G.
Published: July 2009

A new European Project has been launched in April 2008 under the 7th EURATOM Framework Programme (FP7-211333), with a duration of four years, addressing the ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)’. The objective of this project is the development of best practices in the retrieval, treatment and disposal of irradiated graphite & carbonaceous waste-like structural material e.g. non-graphitised carbon bricks and fuel coatings (pyrocarbon, silicon carbide). It addresses both legacy waste as well as waste from future generations of graphite-based nuclear fuel. After defining the various targets for an integrated waste management, comprehensive analysis of the key stages from in-reactor storage to final disposal will then be undertaken with regard to the most economic, environmental and sustainable options. This will be supported by a characterisation programme to localize the contamination in the microstructure of the irradiated graphite and so more to better understand their origin and the release mechanisms during treatment and disposal. It has been discovered that a significant part of the contamination (including 14C) can be removed by thermal, chemical or even microbiological treatment. The feasibility of the associated processes will be experimentally investigated to determine and optimise the decontamination factors. Reuse of the purified material will also be addressed to close the ‘Graphite Cycle’ for future graphite moderated reactors. The disposal behaviour of graphite and carbonaceous wastes and the improvement of suitable waste packages will be another focus of the programme. The CARBOWASTE project is of major importance for the deployment of HTR as each HTR module generates (during a 60 years operational lifetime) about 5,000 to 10,000 metric tonnes of contaminated graphite containing some Peta-Becquerel of radiocarbon. It is strongly recommended to take decommissioning and waste management issues of graphite-moderated reactors already into account when designing new HTR concepts.
Pebble Bed Reactor Dust Production Model
Authors: Cogliati Joshua J. , Ougouag Abderrafi M.
Published: July 2009

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.
Sensitivity Study on Depressurized LOFC Accidents With a Failure of the RCCS in a Modular Gas-Cooled Reactor
Authors: Ha Sang Jun , Kim Seyun , Tak Nam-Il , Lim Hong Sik
Published: July 2009

A modular gas-cooled reactor design with a thermal output of 600MWt and a core exit temperature of 950° C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multi-component problem was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600° C at about 50 hours at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.
Development of the GT-MHR Vertical Turbomachine Design
Authors: Kodochigov N. G. , Belov S. E. , Borovkov M. N. , Shenoy A. , Baxi C. B. , Telengator A.
Published: July 2009

The high-temperature gas-cooled reactor technology is the only nuclear technology capable of achieving coolant temperatures as high as 950 °C and at the same time ensuring safe and efficient production of both electricity and hydrogen. OKBM and GA started independent research in this area in the 1990s. In 1995, OKBM in cooperation with GA started development of the GT-MHR design which combines a safe modular reactor and a power conversion unit based on the high-efficiency Brayton cycle. The power conversion unit in the GT-MHR design has integral configuration, with vertical arrangement of the turbomachine consisting of a synchronous generator and a turbocompressor. Active electromagnetic bearings are used as supports. In order to select optimal technical solutions, the effect of the following factors on the design was considered: vertical or horizontal arrangement, submerged or remote generator with oil bearings, and different turbomachine rotor speeds. Application of electromagnetic bearings and diaphragm coupling between the rotors, integral arrangement of the turbomachine inside the power conversion system vessel, and use of helium as coolant required performance of comprehensive analyses and experiments. For this purpose, the helium turbomachine technology demonstration program was developed and is currently being implemented. This technology demonstration program aims at validating the quantitative and qualitative characteristics of such turbomachine components as electromagnetic and catcher bearings, control system, computer codes, generator, diaphragm coupling, turbocompressor, etc. At the concluding stage of the technology demonstration program, a full-scale turbocompressor model will be tested at a helium test facility. The present paper lists the main parameters of the GT-MHR turbomachine and describes the status of experimental validation of its components.
A Novel Buoyantly-Driven Shutdown Rod Design for Passive Reactivity Control of the PB-AHTR
Authors: Blandford E. D. , Peterson P. F.
Published: July 2009

The modular Pebble Bed Advanced High Temperature Reactor (PB-AHTR), with a nominal power output of 900 MWth, is the most recent US Berkeley design for a liquid fluoride salt cooled reactor. Due to the high volumetric heat capacity of the primary coolant, the PB-AHTR operates with a high power density core with a similar average coolant temperature as in modular helium reactors. The reactivity control system for the PB-AHTR uses a novel buoyantly-driven shutdown rod system that can be actively or passively activated during reactor transients. In addition to a traditional active insertion mechanism, the new shutdown rod system is designed to also operate passively, fulfilling the role of a reserve shutdown system. The physical response of the shutdown rod was simulated both computationally and experimentally, using scaling arguments where applicable, with an emphasis on key phenomena identified by a preliminary PIRT study. This paper discusses preliminary results from this effort.
Numerical Simulations of Jet Flow Mixing Within Rod Bundles
Authors: Salpeter Nathaniel O. , Hassan Yassin A.
Published: July 2009

The goal of the work is to investigate the abilities of the realizable k-ε and the k-ω turbulence models to predict phenomena expected in the lower plenum of the prismatic gas cooled reactor by benchmarking them against one another as well as verifying them against available experimental data [6]. Simulations were performed with CD-Adapco’s Star-CCM+ computational fluid dynamics (CFD) code utilizing the steady state approximation of the realizable k-ε turbulence model and the unsteady RANS approximations of the shear stress transport (SST) k-ω turbulence model. The unsteady results were averaged over a period of time corresponding to one full fluctuation cycle of the phenomena present with the lowest frequency. A case with a single jet of Reynolds number 12,700 was simulated as well as a case with dual jets of Reynolds numbers 6,300 and 12,700. Impingement of the jets occurred on the lower plane of the setup as occurs in the lower plenum. A two-layer shear driven [8] y+ wall treatment was used to satisfy the boundary layer profile. A mesh of 8.6 million polyhedral cells was generated to capture critical flow characteristics within the domain. Polyhedral cells were chosen for this application because they provided a better quality mesh and reduced the total number of cells necessary to achieve accurate results [4]. The simulations carried out were defined as having reached a converged solution when all residuals reduced to less than 10−5. The simulations of the flow in the rod bundle were successful in providing insight into locations of some key recirculation zones and the dependence they have on inlet conditions. The comparison between numerical and experimental results showed similar key flow patterns as well as aided in possible points of focus for future investigation. The information obtained and conclusions drawn will be critical in future numerical benchmarks in this area of research.
Measurements of Flow Modification by Particle Deposition Inside a Packed Bed Using Time-Resolved PIV
Authors: Dominguez-Ontiveros Elvis E. , Estrada-Perez Carlos , Hassan Yassin A.
Published: July 2009

In the Advanced Gas Cooled Pebble Bed Reactors for nuclear power generation, the fuel is spherical coated particles. The energy transfer phenomenon requires detailed understanding of the flow and temperature fields around the spherical fuel pebbles. Detailed information of the complex flow structure within the bed is needed. Generally, for computing the flow through a packed bed reactor or column, the porous media approach is usually used with lumped parameters for hydrodynamic calculations and heat transfer. While this approach can be reasonable for calculating integral flow quantities, it may not provide all the detailed information of the heat transfer and complex flow structure within the bed. The present experimental study presents the full velocity field using particle image velocity technique (PTV) in a conjunction with matched refractive index fluid with the pebbles to achieve optical access. Velocity field measurements are presented delineating the complex flow structure.
Simulation of an X-RAY Fluorescence Experiment for Gold Assaying With Photon Interaction Probability and Moments Model
Authors: Koreshi Zafar U. , Siddiq Sadaf , Khan Irfan A. , Shah Tasneem
Published: July 2009

An interaction probability model is used to model the Energy Dispersive X-Ray Fluorescent (EDXRF) response from a given gold sample to determine the purity (or karatage) of gold. The application of this model is considered for estimating experimental results using a spectrometer with a X-Ray source of 18.6 keV which counts L-shell fluorescence in gold. In the elementary model, Compton scattering is largely ignored as it forms an insignificant contribution compared with photoelectric absorption in the energies of interest. The model is used to infer the “average depth” to which this analysis can be expected to provide reliable results, since ‘depth profiling’ is a major limitation of XRF. This model is further strengthened by carrying out a ‘deep-penetration’ analysis for the energy-dependent photon transport equation with the moments method. The first four moments are computed and used to obtain the spatial profile for the energy-dependent photon flux in the vicinity of L-shell phenomena. This analysis is useful to understand the contribution from interaction phenomena deeper within the material. For validation of depth profile, 2 gold-coated discs of thickness 10 microns and 40 microns were analyzed on the same XRF spectrometer. Transport theory depth-profiling results were validated from detection of the nickel-silver substrate of the discs.
PACE 1450: An Experimental Test Setup for the Investigation of the Crack Behaviour of Prestressed Concrete Containment Walls
Authors: Herrmann Nico , Niklasch Christoph , Kiefer Daniela , Gerlach Lutz , Le Pape Yann , Fortier Ste´phane
Published: July 2009

Among other topics the crack initiation and the leakage through prestressed concrete walls of reactor containments are matters of particular interest for the safety of nuclear facilities. Experimental investigations under different loads which also include the limit state under accidental conditions are the basis for the implementation of these problems in finite element codes. The validity of models that are used in nonlinear calculations has to be verified by comparing their performance with experimental results. In order to determine the capability of simulations to predict the structural behaviour of realistic and representative structural parts these kind of complex tests are indispensable. The PACE 1450 experimental project is an intermediate sized experiment to investigate the behaviour of a curved specimen which is representative for the prestressed containment of a 1450 MWe nuclear power plant. The specimen is loaded by air pressure simulating the internal pressure within the reactor containment under inspection and accidental conditions. The resulting ring tensile stresses of the cylindrical part of the containment are applied externally by eight hydraulic jacks. The initial prestressing of the specimen is realised in such a way that a decreasing of the prestressing force for the purpose of simulating the aging of the structure is possible.
Phosphate Ions: Does Exposure Lead to Degradation of Cementitious Materials?
Authors: Naus Dan J. , Graves Herman L. , Mattus Catherine H. , Dole Leslie R.
Published: July 2009

An assessment of the potential effects of phosphate ions on cementitious materials was made through a review of the literature, contacts with concrete research personnel, and conduct of a “bench-scale” laboratory investigation [1]. The objectives of this limited study were to: (1) review the potential for degradation of cementitious materials due to exposure to high concentrations of phosphate ions; (2) provide an improved understanding of any significant factors that may lead to a requirement to establish exposure limits for concrete structures exposed to soils or ground waters containing high levels of phosphate ions; (3) recommend, as appropriate, whether a limitation on phosphate ion concentration in soils or ground water is required to avoid degradation of concrete structures; and (4) provide a “primer” on factors that can affect the durability of concrete materials and structures in nuclear power plants. Results of a literature review, contacts with industry personnel, and a laboratory investigation indicate that no harmful interactions occur between phosphate ions and cementitious materials unless phosphates are present in form of phosphoric acid. Relative to the “primer,” NUREG/CR-6927, published in February 2007, provides a review of pertinent factors that can affect the durability of nuclear power plant reinforced concrete structures.
Research on the Computerized Symptom-Oriented Procedures of Nuclear Power Plants
Authors: Liu Fei , Zhang Zhijian , Peng Minjun
Published: July 2009

New methods of information presentation and interface design are changing the working conditions in the modern Nuclear Power Plant (NPP) control room. Symptom-oriented EOPs (SOPs) with their structures and practical application are described. The Computerized Symptom-oriented Operating Procedures (CSOP) is researched, which can help the operator analyze all the symptom signals of steam generator tubes rupture (SGTR) and provide the computerized procedures corresponding to the symptom signals. This paper analyzes the accident of SGTR; the accident management of SGTR is important in reactor safety because SGTR is one of the relatively high-frequency events in pressurized water reactors PWRs. The symptom signals of SGTR and the possible accidents corresponding to the symptom signals are analyzed. The homologous measures of symptoms are summarized. The disposal of SGTR adopts the method of based on symptoms. The programs are developed by VxWorks that is a real-time operating system. The debugging of programs is processed on simulator. The test results indicated that the programs can provide operating procedures according to the symptoms of accidents. After adopting the Computerized Emergency Operating Procedures, the labor intensity and mental burden of operators are lightened. Computerized Emergency Operating Procedures can enhance the reliability, safety and efficiency of Nuclear Power Plant.
Analyses of a LBE-Diphyl THT Heat Exchanger With the System Code Trace
Authors: Jaeger Wadim , Sa´chez Espinoza Victor H.
Published: July 2009

The aim of this investigation was the application of the TRACE code on a complex heat exchanger (HX) design to qualify the TRACE capabilities to simulate liquid metals as coolants. The chosen HX is related to the MEGAPIE project. The liquid LBE (lead-bismuth-eutectics) cooled target is coupled via a HX to Diphyl THT (DTHT), a diathermic fluid. At the CHEOPE test facility of ENEA in Brasimone (Italy) a mock-up of this HX has been build to investigate the heat transfer behavior between the LBE and the DTHT. Several experiments have been performed at the CHEOPE test facility. The experimental setup consists of one loop for the LBE, one for the DTHT, a HX and of a heater section which represents the target area. For this study only the HX was modeled and investigated. The experimental values for the HX entrance temperatures and mass flow rates served as boundary condition. One challenge of this post-test investigation was to model the DTHT loop because the innately TRACE version can not handle this fluid. Hence, the first step was to implement the thermal properties of DTHT into the TRACE source code. First calculated results obtained with the original TRACE version were disappointing. TRACE was not able to reproduce the HX exit temperatures. Consequently appropriate correlations for helical channel, derived from experimental investigations or CFD analyses, were implemented into TRACE. The results of the modified TRACE version were much closer to the experiments than those for the original version.
IMTHUA Methods for Thermal-Hydraulics Code Structure Uncertainty Assessment
Authors: Pourgol-Mohammad Mohammad , Mosleh Ali , Modarres Mohammad
Published: July 2009

A successful treatment of model uncertainty results in an expression of uncertainty that includes the true value at some stated level of confidence. Code structure uncertainties (model uncertainty) are a crucial source of uncertainty quantification for thermal-hydraulics (TH) system codes such as RELAP5, TRAC, and recently consolidated TRACE code. These codes are an assembly of models and correlations for simulation of physical phenomena and behavior of system component. In some cases there are alternative sub-models, or several different correlations for calculation of a single phenomenon of interest. There are also “user options” for choosing one of several models or correlations in performing a specific code computation. Dynamic characteristics of TH add more complexity to the code calculation, meaning for example, that specific code models and correlations invoked are sequence-dependent, and based certain (dynamic) conditions being satisfied. This paper discusses the techniques developed in the Integrated Methodology for Thermal-Hydraulics Uncertainty Analysis (IMTHUA), specifically for the treatment of uncertainties due to code structure and models. The methodology comprehensively covers various aspects of complex code uncertainty assessments for important accident transients. It considers the TH code structural uncertainties explicitly by treating internal sub-model uncertainties, and by propagating such model uncertainties in the code calculations (including uncertainties about input parameters.) Structural uncertainty assessment (model uncertainty) for a single model will be discussed by considering “correction factor”, “bias”, and also through sub-model output updating with available experimental evidence. In case of multiple alternative models, techniques of dynamic model switching, user controlled model selection, model mixing, and model maximization/minimization will be discussed. Examples from different applications including, Marviken test facility blowdown, LOFT LBLOCA and a typical PWR LOCA scenario calculations will be provided for greater clarification of the proposed techniques.
On Advanced Fuel With Improved Properties
Authors: Kolokol Alexander S. , Shimkevich Alexander L.
Published: July 2009

A disadvantage of the uranium dioxide fuel is very low thermal conductivity than the one of nitride, carbide, metal fuel, and cermets as composites, UO2+Me, due to the portion in thermal conductivity of their electronic conductivity and high phonon mobility. An investigation of the microstructure and atomic dynamics of solid solutions as well as the physical and chemical processes in them will make it possible to adjust the properties of the solutions in steps according to prescribed indicators by using alloying additives. The concept for designing an oxide fuel may be promising for the development of a new generation of nuclear reactors. In developing the methods for designing reactor materials as to the nuclear fuel, microscopic-structure improving its thermal and physical properties is formulated here.
Methodology for Evaluating Radioactive Levels of Material to Be Released Under Newly Introduced Clearance System in Japan
Authors: Kishioka Kazuhiko
Published: July 2009

The amendment of nuclear reactors regulation law was established and the clearance system was introduced in 2005 in Japan.
Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor
Authors: Zhang Dalin , Liu Changliang , Qian Libo , Su Guanghui , Qiu Suizheng
Published: July 2009

The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.
Optimizing Project Prioritization Under Budget Uncertainty
Authors: Koc¸ Ali , Morton David , Popova Elmira , Hess Stephen , Kee Ernie , Richards Drew
Published: July 2009

We consider a problem commonly faced in the nuclear power industry, involving annual selection of plant capital investments under the constraints of a limited and uncertain budget. When the budget is assumed known, a typical approach to such problems is built on a multi-dimensional knapsack model. This model takes as input the available budget in each year, the stream of liabilities induced by selecting each project, and the profit, i.e., net present value (NPV), of each project. The goal is to select the portfolio of projects with the highest total NPV, while observing the budget constraint for each year, as well as any additional constraints. We show that a portfolio selected in this manner can fail to hedge against uncertainties in the budget. While the budget may be known at the beginning of the planning period, external events can cause this to change as time unfolds, and hence the funds that will actually be allocated over time are typically uncertain. So, we propose a model that forms an optimal priority list of projects, incorporating multiple budget scenarios. The model is applied to example projects from the South Texas Project Nuclear Operating Company (STPNOC).
Experiments on Air-Water Countercurrent Flow in a Rectangular Duct Simulating PWR Hot Leg
Authors: Minami Noritoshi , Nishiwaki Daisuke , Kataoka Hironobu , Tomiyama Akio , Hosokawa Shigeo , Murase Michio
Published: July 2009

In the case of loss of the residual heat removal system under mid-loop operation during shutdown of the pressurized water reactor (PWR) plant, steam generated in a reactor core and condensed water in a steam generator (SG) form a countercurrent flow in a hot leg. In this study, in order to improve a counter-current flow model of a transient analysis code, experiments were conducted using a scale-down model of the PWR hot leg, and flow patterns and counter-current flow limitation (CCFL) characteristics were measured. A rectangular duct, whose height is about 1/5th of the hot leg diameter, was used to simulate the hot leg, and air and water at atmospheric pressure and room temperature were used for gas and liquid phases. In the horizontal section, as air flow rate QG increases, the flow pattern transits from a stratified flow to wavy flow, and then wavy to wavy-mist flow. When the latter transition takes place, water flow from the horizontal duct to the lower tank is to be restricted. Flow patterns in the elbow section are the same as those in the horizontal section. Wavy flow is not formed in the inclined section, where the transition to wavy-mist flow occurs due to the inflow of wavy-mist flow generated in the horizontal section. Flow patterns in the elbow and inclined section are strongly affected by those in the horizontal section. CCFL characteristics are well correlated with the Wallis-type correlation, and the onset of CCFL well corresponds to the transition from wavy flow to wavy-mist flow.
Experimental Investigation and CFD Simulation of the Behaviour of Mineral Wool in the Reactor Sump
Authors: Krepper Eckhard , Cartland-Glover Gregory , Grahn Alexander , Weiss Frank-Peter , Alt So¨ren , Hampel Rainer , Ka¨stner Wolfgang , Seeliger Andre´
Published: July 2009

The investigation of insulation debris generation, transport and sedimentation becomes important with regard to reactor safety research for PWR and BWR, when considering the long-term behavior of emergency core cooling systems during all types of loss of coolant accidents (LOCA). The insulation debris released near the break during a LOCA incident consists of a mixture of disparate particle population that varies with size, shape, consistency and other properties. Some fractions of the released insulation debris can be transported into the reactor sump, where it may perturb/impinge on the emergency core cooling systems. Open questions of generic interest are the sedimentation of the insulation debris in a water pool, its possible re-suspension and transport in the sump water flow and the particle load on strainers and corresponding pressure drop. A joint research project on such questions is being performed in cooperation between the University of Applied Sciences Zittau/Go¨rlitz and the Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description. While the experiments are performed at the University at Zittau/Go¨rlitz, the theoretical modeling efforts are concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modeling are described and feasibility studies including the conceptual design of the experiments are presented.
Treatment of Epistemic and Aleatory Uncertainties in the Statistical Analysis of the Neutronic Protection System in CANDU Reactors
Authors: Novog D. R. , Atkinson K. , Levine M. , Nainer O. , Phan B.
Published: July 2009

In existing and new nuclear power plants a variety of special safety systems are employed which will trigger fast reactor shutdown in the event of an accident or undesirable plant condition. A key consideration in determining the optimal trip setpoints for these safety systems is the treatment of both epistemic and aleatory uncertainties. A significant issue also arises in attempting to construct probabilistic methodologies that accurately account for these uncertainties while still maintaining consistency with instrument uncertainty calculation methodologies like ISA 67.04. Furthermore, for probabilistic based analyses such as Loss of Regulation events in a CANDU reactor, the calculation of the neutron overpower trip setpoint involves extremal functions and as such extreme value statistics are applied. Since there are variations in the actual reactor physics parameters (fuel channel power, flux detector drift and total reactor power) and thermalhydraulic conditions, determination of accident behavior is more complicated than if the reactor conditions were completely fixed in time. This paper presents the methodology used to establish the epistemic and aleatory uncertainties in reactor trip setpoints such that margins to safety can be established and so that quantitative statistical statements can be made on the probability of safety system action and the confidence level which is self-consistent with the instrument calculation methodology outlined in ISA 67.04. This paper examines the trip setpoints required to mitigate Loss of Power Regulation accidents in a CANDU reactor using a unique treatment of the epistemic and aleatory uncertainties. In addition to the statistical treatment, a key finding in this work relates to the specification of the initial reactor state based on synchronous plant measurements as opposed to Monte Carlo generated states based on sampling of fitted probability distributions.
Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel
Authors: Baily Carl E. , Moore Karen A. , Knight Collin J. , Wells Peter B. , Petersen Paul J. , Siahpush Ali S. , Weseman Matthew T.
Published: July 2009

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.
A Calibration of a Feedwater Flowmeter Installed to Model Loop of a Nuclear Power Plant Using High Reynolds Number Facility
Authors: Furuichi Noriyuki , Terao Yoshiya , Takamoto Masaki
Published: July 2009

A calibration result of ultrasonic flowmeters using in a feedwater flowrate in a nuclear power plant, is described under a variety of upstream conditions using the new high Reynolds number calibration facility. The pipe layouts are classified to five type three-dimensional one with two or three elbows. The flow conditioners are tube bundle type and Mitsubishi type. Pipe Reynolds number is up to 1.6×107. The large effect of the flow conditioner and pipe layout is observed for cramp-on type. For multi-path type, individuality is observed.
Effect of a Dimple Location in a Spacer Grid on the Impact Strength
Authors: Song Kee-Nam , Lee Soo-Bum
Published: July 2009

A spacer grid is one of the most important structural components in a PWR fuel assembly. The spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is an interconnected array of slotted grid straps welded at the intersections to form an egg-crate structure. Dimples and springs are stamped into each grid strap to support the fuel rods. Zircaloy is prevailing as the material of the spacer grid because of its low neutron absorption characteristic and its extensive successful in-reactor use. The primary considerations are to provide a Zircaloy spacer grid with crush strength sufficient to resist design basis loads, without significantly increasing pressure drop across the reactor core. Generally, a Zircaloy grid strap’s thickness and average height have been the main design variables in order to meet the above considerations. Recently, it was reported that a dimple location is also a design variable that affects the crush strength of a spacer grid. In this paper, based on this fact, the effect of a dimple location in a 3×3 sub-size support grid on the impact strength has been investigated as a preliminary parameter study for a 16×16 full size support grid. A 3×3 model is chosen to facilitate in a fast computation for an optimization. The optimal location of the dimple has been found and some design guidelines for a support grid have been deduced in this work.
Regulatory Guide 1.200 PRA Peer Reviews: A Peer Reviewer Perspective
Authors: Finnicum David J.
Published: July 2009

In January 2007, the NRC issued Revision 1 to Regulatory Guide (RG) 1.200 endorsing Addendum B of the ASME PRA Standard, RA-Sb-2005. Effective January 1, 2008, the probabilistic risk assessment (PRA) supporting any risk-informed application must comply with the requirements of RG 1.200 and the ASME PRA Standard, as explained in NRC’s Regulatory Issue Summary (RIS) 2007-06. Per Section 6 of the ASME PRA Standard, a peer review must be performed to demonstrate compliance with the ASME PRA Standard. In the late 1990s, all nuclear plants in the U. S. had a peer review performed in accordance with NEI 00-02. However, the ASME PRA Standard requirements are more extensive than the review elements in NEI 00-02. Appendix D of NEI 00-02, endorsed by Appendix B of RG 1.200, includes a self-assessment process for using the NEI 00-02 results to, in part; assess a PRA’s compliance with the ASME PRA Standard. The ASME PRA Standard requires a peer review for any PRA element that has been upgraded since the original peer review. There is also the consideration that if the PRA updates have resulted in enough changes to the model that the current model no longer is adequately represented by the model that was originally peer reviewed, then, at their discretion, the utility may request a new peer review. Both the Boiling Water Reactor Owners Group (BWROG) and the Pressurized Water Reactor Owners Group (PWROG) have authorized projects to conduct the peer reviews against RG 1.200, Revision 1 to support risk-informed applications after January, 2008. Two key elements have changed since the original NEI 00-02 peer reviews. These are: (1) the number of specific requirements that need to be reviewed and (2) the need for the utility undergoing a peer review to complete a self-assessment and a roadmap to the documentation supporting their assessment to enable the peer review team to complete their review within a reasonable timeframe. Since the original approval of the ASME PRA Standard in 2003, many plants have performed self-assessments either internally or with the assistance of contractors to ascertain their compliance with the ASME PRA Standard. A number of lessons have been learned as a result of these self-assessments that should be carried forward to the next round of formal peer reviews against Addendum B of the ASME PRA, RA-Sb-2005. This paper examines some of the lessons learned while performing these self-assessments, and provides some guidance for a pre-review self-assessment and as well as some guidelines on compiling a roadmap to support the self-assessment and the peer review.
Subcooled Boiling Heat Transfer for Turbulent Flow of Water in a Short Vertical Tube
Authors: Hata Koichi , Noda Nobuaki
Published: July 2009

The subcooled boiling heat transfer for Platinum test tube divided into three sections (upper, mid and lower positions) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet liquid temperatures (Tin = 295.26 to 305.25 K), the inlet pressures (Pin = 739.26 to 1064.48 kPa) and the exponentially increasing heat input with various periods (Q = Q0 exp(t/τ), τ = 22.52 ms to 26.31 s) was systematically measured by an experimental water loop comprised of a pressurizer. The Platinum test tube of inner diameter (d = 3 mm), heated length (L = 66.5 mm), L/d (= 22.17) and wall thickness (δ = 0.5 mm) with a commercial finish of inner surface (average roughness, Ra = 0.40 μm) is used in this work. The outer surface temperature of the test tube was observed by an infrared thermal imaging camera. The axial variations of the inner surface temperature, the heat flux and the heat transfer coefficient from non-boiling to CHF were clarified. The subcooled boiling heat transfer for Platinum test tube with a commercial finish of inner surface was compared with the values calculated by other workers’ correlations for the subcooled boiling heat transfer. The influence of exponential period (τ) and flow velocity (u) on the subcooled boiling heat transfer is investigated into details and the predictable correlation of the subcooled boiling heat transfer for turbulent flow of water in a short vertical tube is derived based on the experimental data for Platinum test tube with a commercial finish of inner surface. The correlation can describe the subcooled boiling heat transfer coefficients obtained in this work within 15% difference.
Assessment of Debris Bed Formation Characteristics Following Core Melt Down Scenario With Simulant System
Authors: Harvey J. , Narayanan K. S. , Das S. K. , Rao E. V. H. M. , Lydia G. , Malarvizhi B. , Murthy S. S. , Kumaresan M. , Kasinathan N. , Rajan M.
Published: July 2009

In a fast reactor safety analysis determination of the molten core conditions when it reaches the core catcher plate is one of the main factors after a postulated MFCI event. If large fragmentation and quenching is accomplished in the coolant column no major problems for main vessel attack would occur. If instead, a significant amount of melt would remain as a solid molten cake, potential for lower head penetration would exist. In the present study towards development of a model for core melting and debris settling on to a core catcher plate, early phase of liquid stream fragmentation progression due to hydrodynamic consideration was investigated with woods metal melt water system. The system selected simulates the hydrodynamic physical properties more closely that of liquid UO2-sodium system. Assessment of debris-bed forming characteristics was carried out with different coolant column and different melt temperatures with melt inventories up to 20 kg released from a nozzle of 8 mm diameter. Bed height, debris spread area, jet break up length and repose angle obtained are presented for a melt release rate of ∼ 600 g/s. Only solidified debris constituted the bed for a melt temperature of 100°C and water temperature of 29 °C, with 720 mm water height. The estimated average bed height, bed porosity and heap repose angle were 15 cm, ∼0.6 and 43° respectively. Solidified central columnar lump of height 30 cm was seen for a water column of 360 mm. Relative bed forming characteristics for melt temperatures of 120 °C and 220 °C are also presented. High speed video imaging was taken to assess the stream break up distance and heap formation dynamics. Bulk coolant temperatures close to the melt stream were also monitored. Dependence of particulate debris and bed characteristics on melt temperature, interaction height and melt inventory have been brought out.