bluuebunny/crossref_metadata_2025 huggingface dataset:
Generalized Tolerance Limit Evaluation Method to Determine Statistically Meaningful Minimum Code Simulations
Authors: Hong In Seob , Connolly Adrian
Published: July 2009

The tolerance limit approach has been internationally adopted for nuclear safety analyses as a way of identifying more realistic safety limits in the Best Estimate Plus Uncertainty (BEPU) methodology. The approach is based on the Wilks’ formula set suggested by the GRS (Gesellschaft fu¨r Anlagen-und Reaktorsicherheit), and consists of two approaches; one-sided approach to obtain a statistical meaning of the maximum output value in the multiple code runs, and two-sided approach for the minimum and maximum values at the same time. The Wilks’ formula set suggested by the GRS has been employed without questioning any ambiguities in its true meaning, especially in the two-sided approach. In this paper, the formula set is derived based on the binomial distribution as a direct vehicle to the formulas for clarification purposes. This approach is distinguished from the present integral form of the original Wilks’ formula in that it is based on the probability density function and the joint probability density function associated with the order statistics. The purposes of this paper are to derive the true meanings of the present Wilks’ formulas, which are applied in the nuclear safety analysis, and to derive a new set of formulas to determine statistically meaningful minimum number of code runs in the BEPU. An improved formula set, which consists of the Generalized Maximum Tolerance Limit (GMTL) and the Generalized Max-Mini Tolerance Limit (GMMTL), is proposed with an emphasis on a new form of the two-sided formula: Centered Two-sided Tolerance Limit (CTTL) formula, which is directly derived from the GMMTL formula. For example, a minimum number of 146 code runs is suggested in lieu of the present 93 code runs to achieve the 95th percentile with 95% confidence in the two-sided approach.
Second Generation Experimental Equipment Design to Support Voloxidation Testing at INL
Authors: Wahlquist Dennis L. , Bateman Kenneth J. , Westphal Brian R.
Published: July 2009

Voloxidation is a potential process used to prepare fuel for spent-oxide-fuel treatment. The spent oxide fuel is heated to an elevated temperature in oxygen or air to promote separation of the fuel from the cladding. Low pressures are also required to enhance volatizing the fission products. The Idaho National Laboratory and the Korea Atomic Energy Research Institute have been collaborating on voloxidation research through a joint International Nuclear Energy Research Initiative project. The new high temperature multiple zone furnace and off-gas trapping system (OTS) are designed for remote operation in the Hot Fuel Examination Facility. The OTS is designed to handle vacuum, oxidizing environments, and high temperatures. Distinctive temperature zones of the furnace are used to selectively capture the fission products. Vacuum conditions at elevated temperatures are achieved by incorporating various metal seals. The OTS has proved durable under these aggressive operating conditions. A detailed description of the second generation furnace and OTS with enhanced capabilities will be presented.
Low Boron Level Loading Pattern
Authors: Gu¨ler Cenk , Brown Jeffery A.
Published: July 2009

A new core loading pattern scheme has been developed with the objective of significantly reducing both the maximum and cycle average RCS (Reactor Cooling System) soluble boron requirements for a typical PWR (Pressurized Water Reactor) plant with 18 month cycle. This strategy has several advantages including reduced operational costs due to boron addition and subsequent dilution, less challenging coolant chemistry adjustments and decreased tritium production over the cycles. The core design utilizes a typical Westinghouse 4-Loop core with assemblies containing both ZrB2 Integral Fuel burnable Absorber (IFBA) and high concentration Gadolinium (Gd) burnable absorbers (BAs) in the same assembly. This combination achieves the boron reduction objectives while still maintaining good power margins and good fuel cycle economics. This paper describes the major characteristics of these cycles and compares these to a reference cycle design typical of those used in current Westinghouse 4-Loop type plants.
Finding the Minimun of the Quadratic Functional in Variational Approach in Transport Theory Problems
Authors: Panta Pazos Rube´n
Published: July 2009

In this work it is reviewed the variational approach for some Transport Problems. Let X be a convex domain in Rn, and V a compact set. For that, it is considered the following equation: ∂ψ∂t(x,v,t)+v·∇ψ(x,v,t)+h(x,μ)ψ(x,v,t)==∫Vk(x,v,v′)ψ(x,v′,t)dv′+q(x,v,t)(1) where x represents the spatial variable in a domain D, v an element of a compact set V, Ψ is the angular flux, h(x,v) the collision frequency, k(x,v,v’) the scattering kernel function and q(x,v) the source function. It is put the attention in the construction of the quadratic functional J which appears in variational approaches for transport theory (for example, the Vladimirov functional). Some properties of this functional in a proper functional framework, in order to determine the minimum for J are considered. First, the general formulation is studied. Then an algorithm is given for minimizing the functional J for two remarkable problems: spherical harmonic method and spectral collocation method. A program associated to this algorithm is worked in a computer algebraic system, and also was depeloped a version in a high level language.
Refined Modeling of Projectile Impact Onto Submerged Structure
Authors: Onisoru Justin , Coman Ovidiu , Wilson Paul , Thomas George
Published: July 2009

Structural integrity of spent fuel racks is a critical safety issue in nuclear power stations. The standard approach of evaluating the effects of an impact projectile on a submerged structure, which constitute the start point of the current study, involves three main steps: determination of the conditions just prior to the impact (that are considered as initial conditions for the analysis), setting the mechanism of transferring energy from the projectile to the target structure, and determining how that energy is absorbed by the impacted structure. Usually, the dynamics of the projectile are ideally considered, the influence of the fluid presence is restricted to the determination of the impact velocity and strain rate dependency is limited to choosing a true stress vs. strain curve corresponding to some constant strain rate. Starting from the standard engineering approach, the authors have refined the model considering more realistic dynamics of the projectile, extending the influence of the fluid to the entire analysis and using a more accurate strain rate dependant material behavior. Explicit Finite Element analyses are used in order to incorporate the desired effects.
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: 6 — Study of Eutectic Reaction Between Metals: FPMD Approach
Authors: Himi Masashi , Kozakai Hiroshi , Yamamoto Yuichi , Hosoda Seigo , Shirakawa Noriyuki , Arima Tatsumi
Published: July 2009

A five-year research project started in FY2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with MPS for Reactor Safety Analysis) [1]. In this project, both mixed-oxide (MOX) and metal fuels are considered as a fuel material component. One of the main features of the project is to investigate eutectic reactions between the metal fuel and the cladding/structure materials with phase diagram calculation, classical and first principle molecular dynamics (CMD and FPMD) methods. This paper presents outcomes from study with FPMD.
Development of Low-Activation Design Method for Reduction of Radioactive Waste Below Clearance Level
Authors: Kimura Ken-Ichi , Hasegawa Akira , Hayashi Katsumi , Uematsu Mikio , Ogata Tomohiro , Tanosaki Takao , Yoshino Ryoetsu , Sato Mituru , Saito Minoru , Kinno Masaharu
Published: July 2009

Design methodology for reinforced concrete of nuclear power plants to reduce radioactive wastes in decommission phase has been developed. To realize this purpose, (1) development of raw materials database of cements, aggregates and steel bars on concentration of radioactive target elements, (2) trial production of low activation cements and steel bars based on the material database developed in (1), and (3) development of tools for estimation and prediction of the amount of radioactive elements in reactor shielding walls have been carried out. Radioactive analysis showed that Co and Eu were the major target elements which decide the radioactivity level of reinforced concrete from wide survey of raw materials for concrete (typically aggregates and cements). Material database for the contents of Co and Eu was developed based on the chemical analysis and radioactivation analysis. Upon the above survey and execution expreiment of concrete, six types of low-activation concrete are proposed for various radioactive portion in the plant. These concrete have a 1/10 – 1/300 rasioactivity compare to the ordinary concrete, which are assumed the concrete with Andesite aggregate and ordinary Portland cement. Baed on the above data base, it was clarified that the low activation cement would be successfully manufactured by adequate selection of raw materials. The prospect to produce the low-heat portland cement which would have a 1/3 radioactivity in comparison with conventioanl cements obtained by means of selection of limestone and natural gypsum. An attempte was carried out to produece low activation heavy-mortar which would have radioactivity below the clearance level when using at the radiation shielding wall of BWR. Characterization and optimization of consturction conditions with new additives have also been carried out. These two new raw materials for low-activation concrete are conducted in pre-manufacture size, and over the laboratry level. Boron added low-activation concrete are also carried out as extreamly high performance low-activation concrete. It was claryfied that the accurcy of calculation results of the radioactivity evaluation was very high compared to available benchmark calculation for the JPDR and commercial light water reactor. The specification of the mapping system for judging the activation classification was also developed by using the general-purpose radio activation calculation tool. This work is supported by a grant-in-aid of Innovative and Viable Nuclear Technology (IVNET) development project of Ministry of Economy, Trade and Industry, Japan.
Single Phase Pressure-Drop Tests in 7-Element Bundles With Different Endplug Shapes
Authors: Yang S. K.
Published: July 2009

Pressure-drop characteristics in CANDU® fuel bundles are of importance for critical power analysis and safety trip analysis for the reactor core. Different kinds of endplug shapes are being used in different fuel designs. The difference in endplug shape may result in slightly different hydraulic characteristics of the fuel bundles. Single-phase pressure-drop tests were performed in aligned and misaligned (15° and 30°) 7-element bundle strings with three different endplug shapes (Type 1, 2 and 3): Type 1 and 3 with round surface, and Type 2 with flat surface. The test results showed that the bundle-loss coefficients of the misaligned bundle with Type 2 having flat surface presented the highest value among the three-endplug types: 8.6% higher on average, compared to those of the misaligned bundle with Type 1 endplugs (Type 1 as a reference).
Information Entropy: Managing Unknown Risks, Disasters and Losses
Authors: Duffey Romney B. , Saull John W.
Published: July 2009

We establish the probability and risk for the next disaster or loss outcome. We select proactive indicators of progress to support risk management and cultural analysis, and hence make predictions of the next outcome probability based on data. We answer the questions asked by Bernoulli and Bayes so long ago: knowing what we know has already happened, how we can predict what will happen next. With limited prior knowledge, the fundamental paradox is that we cannot learn other than by making those very errors, which we seek to avoid. We show how we must follow the Universal Learning Curve, even if we have had no outcomes and may face an unknown risk. Assuming that the future probability, rate and number are based solely on the past experience, we have shown formally the prediction of the future outcome probability and number. Future rate and number reductions will depend on our sustaining a true learning environment. For rare events and risks, we also determine the probability of having (or observing) the very next unknown. We show that it does not matter if we do not know the exact numbers: the learning trend is the key for decision making and risk taking. The rational choice and implication is to trust our experience and not to be afraid of the risk due to the perceived unknown.
Characterization and Structural Integrity Tests of Ex-Service Steam Generator Tubes
Authors: Pagan Sandra , Duan Xinjian , Kozluk Michael J. , Mills Brian , Goszczynski Guylaine
Published: July 2009

Canadian nuclear standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation in mechanical properties of Monel 400 tubes after 7 to 18 Effective Full Power Years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.
Detached Eddy Simulations and Transient RANS Simulations of Turbulent Flow in the Lower Plenum of a Gas-Cooled Reactor
Authors: Das Kaushik , Basu Debashis , Painter Scott L. , Howard Lane D. , Green Steven T.
Published: July 2009

This paper presents preliminary results of a computational study conducted to analyze the turbulent flow in the lower plenum of an advanced next generation gas-cooled reactor. The turbulence models used in the current simulations are the Detached Eddy Simulation (DES) model and the transient RANS (Reynolds Averaged Navier Stokes) model. The current study is limited to flow in a row of confined cylinders designed to mimic a model of a prismatic gas-cooled reactor lower plenum design. The experimental configuration consists of a finite array of short graphite posts supporting the reactor core. Five cylinders, which represent vertical support posts in the lower plenum of an advanced reactor concept, are emplaced on the cross-stream centerline. In the current work, an idealized model was used to model a region of the lower plenum for a simplified set of conditions that enabled the flow to be treated as an isothermal, incompressible fluid with constant properties. The simulated results are compared with available experimental data, which were obtained using three-dimensional Particle Image Velocimetry (PIV). The two-equation realizable k-ε model is used as the baseline model for both the Unsteady Reynolds Averaged Navier Stokes Equations (URANS) as well as the DES simulations. The flow unsteadiness accounts for the fluctuations due to unsteady vortex shedding. The DES simulations predicted the flow unsteadiness more accurately than the URANS simulations. The simulated time-averaged quantities were also compared with the experimental data. The RANS simulations and the DES simulations provide almost same predictions for the time averaged quantities. The predicted results show discrepancies with the experimental results.
Thermalhydraulic Analysis of the Steam Generator Design for the 700 MWe Indian Nuclear Power Plant
Authors: John Benny , Pietralik John M.
Published: July 2009

The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This development involves redesigning and up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, comparing to the current generation of the 540 MWe IPHWRs. The SG used is a vertical, inverted U-tube, natural-circulation type. The 700 MWe SG is of the same type, has the same tube bundle design and the same heat transfer area as the 540 MWe SG. In addition, the tube diameter, tube pitch, and outer diameter of the SG sections are the same as those for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer, and the flow distribution plate are different in the two designs. The changes were required due to the 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis of the SG using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated for global parameters against the 540 MWe SG operating data. The analysis was made for two design versions, (1) with and (2) without integral preheater and for two versions of inlet primary water conditions, (1) with and (2) without primary steam at the entrance to the SG tubes. In addition, the effect of fouling and tube plugging on the long-term operation of the SG was considered. The paper covers the results obtained and discussion on the above-mentioned issues.
Analysis of Two-Phase Flow Instability for Steam Generator
Authors: Hou Su-Xia , Tai Yun , Zhao Fu-Yu
Published: July 2009

Two-phase flow instability is an important problem that affects the running of steam generators in nuclear reactor systems. In this paper, two-phase flow instability in parallel channels of a steam generator are analyzed to disclose the mechanism of flow instability by using the frequency domain method. The mathematical expressions of heat transfer and flow for a steam generator are proposed, and the transfer function of the closed-loop system is deduced by using linearization and Laplace transfer. The steam generator’s stability is judged according to Nyquist stability criterion. Depending on this fundamental principal, the computer code is developed to analyze the stability of steam generators. The results displayed two conclusions; firstly, the increase of inlet orifices or mass flow rate enhances the stability of generator steam; secondly, the coupling interactions between channels and their external loop effects on the stability of generator steam are not ignored. The result show that the effects are non-monotonic on the stability of generator steam.
3-D Turbulent Modeling of Fluid Flow in Primary Side of CANDU Steam Generators
Authors: Ghafouri-Azar Reza
Published: July 2009

Turbulent flow of heavy water through the primary loop of CANDU steam generators is associated with a number of known failure mechanisms such as flow accelerated corrosion, flow induced vibration and mechanical fatigue. The primary causes of these failures, or degradations, are interactions of the turbulence effect of fluid with the surrounding structure. Using Computational Fluid Dynamics and heat transfer techniques the flow fields inside hot leg and cold side of the steam generator were analyzed to study the nature of the flow fields inside the 2 halves of the primary head. The model used to investigate the flow field was a three-dimensional k-omega turbulence flow model. The solutions of the analysis can be used to investigate the effects of flow on components close to the divider plate and tubesheet including locking tabs and skin fix inside the halves. The effect of turbulence was investigated by varying the inlet nozzle sizes and locations.
Modeling Atmospheric Releases From LLRW Mechanical Shredder
Authors: Ledoux Mark , Gardner Jeff , Sandquist Gary M.
Published: July 2009

Mechanical shredders are used on thin metal, large volume, LLRW materials to reduce volume and enhance compaction for disposal operations. However, during shredding, atmospheric releases from the intense mechanical action may be produced and must be monitored and controlled to satisfy regulatory and ALARA emission control requirements. Standard analytical and EPA computer models are explored and experimentally reviewed to determine potential releases given adequate knowledge of input parameters such as radioactive material sources, waste material-volume-configurations, prevailing meteorological conditions, and airborne material characteristics. Results of these studies, modeling efforts, and measurements are presented together with EPA SCREEN3 computer output files.
Tetrahedral-Chain-Cluster Model for Thermodynamic Description of Fluids
Authors: Shimkevich Alexander L.
Published: July 2009

The topological structure of density fluctuations of the condensed matter in different aggregative states (liquid, crystal, and amorphous body) is represented as instant densely packed fours of atoms as tetrahedrons connected in pairs by faces in Bernal’s n-chains. Tetrahedral clusters of the dense part of the matter are investigated within the framework of matrix algebra: the “genetic” structure of Bernal’s n-chains is decoded; for calculating the partition function and configuration entropy of the isotropic matter, the number of distinguishable tetrahedral chains is determined as a function of the number of tetrahedrons in them.
Transmutation Efficiency of Plutonium and Minor Actinides in PHWR
Authors: Balas (Ghizdeanu) Nineta , Ghitescu Petre
Published: July 2009

PHWRs use natural uranium as fuel and consequently the burn-up coefficient is relatively small compared to PWRs or other existing power reactors. The small burn-up coefficient results in a high volume of irradiated fuel to be disposed, with a high concentration of plutonium and minor actinides. In Romania the irradiated fuel from the existing CANDU 6 spent fuel pool is currently transferred in the Dry Intermediate Fuel Storage Facility existing at the NPP site. Partitioning and Transmutation (P&T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. The use for this purpose of ADS and FBR was more studied, but HWR were not. Therefore, the paper presents different theoretical possibilities to transmute/burn the Plutonium and minor actinides in two different PHWRs — CANDU and ACR, using WIMSD code. Different types of MOX alternative fuel, with variable initial Pu content are analyzed. The results present the reactivity effects along with the isotopes concentration in spent alternative fuel and determine the optimal solution for the fuel type/composition. Thus is indicated the most suitable PHWR type of reactor for possible Plutonium and minor actinides transmutation. The simulations showed that Pu content for an irradiation period of 200 days decreases from the initial value up to 11% in a CANDU reactor and 29% in an ACR. Thus ACR can reduce the plutonium inventory from MOX fuel and could be a transmutation solution. From the economic/technical point of view this analysis also provides input for a study yet to be conducted.
An Initial Study on Modeling the United States Thermal Fuel Cycle Mass Flow Using Vensim
Authors: Brinton Samuel , Tokuhiro Akira
Published: July 2009

According to current forecasts, nuclear power plant construction and nuclear-generated electricity production is projected to increase in the next half-century. This is likely due to the fact that nuclear energy is an ‘environmental alternative’ to fossil fuel plants that emit greenhouse gases (GHG). Nuclear power also has a much higher energy density output than other alternative energy sources such as solar, wind, and biomass energies. There is also growing consensus that processing of low- and high-level waste, LLW and HLW respectively, is a political issue rather than a technical challenge. Prudent implementation of a closed fuel cycle not only curbs build-up of GHGs, but can equally mitigate the need to store nuclear used fuel. The Global Nuclear Energy Partnership (GNEP) is promoting gradual integration of fuel reprocessing, and deployment of fast reactors (FRs) into the global fleet for long-term uranium resource usage. The use of mixed oxide (MOX) fuel burning Light Water Reactors (LWR) has also been suggested by fuel cycle researchers. This study concentrated on modeling the construction and decommissioning rates of six major facilities comprising the nuclear fuel cycle, as follows: (1) current LWRs decommissioned at 60-years service life, (2) new LWRs burning MOX fuel, (3) new (Gen’ III+) LWRs to replace units and/or be added to the fleet, (4) new FRs to be added to the fleet, (5) new reprocessing and MOX fuel fabrication facilities and (6) new LWR fuel fabrication facilities. Our initial work [1] focused on modeling the construction and decommissioning rates of reactors to be deployed. This is being followed with a ‘mass flow model’, starting from uranium ore and following it to spent forms. The visual dynamic modeling program Vensim was used to create a system of equations and variables to track the mass flows from enrichment, fabrication, burn-up, and the back-end of the fuel cycle. Sensible construction and deployment rates were benchmarked against recent reports and then plausible scenarios considered parametrically. The timeline starts in 2007 and extends in a preliminary model to 2057; a further mass flow model scenario continues until 2107. The scenarios considered provide estimates of the uranium ore requirements, quantities of LLW and HLW production, and waste storage volume needs. The results of this study suggest the number of reprocessing facilities necessary to stabilize and/or reduce recently reported levels of spent fuel inventory. Preliminary results indicate that the entire national spent fuel inventory produced over the next ∼50 years can be reprocessed by a reprocessing plant construction rate of less than 0.07 plants/year (small capacity) or less than 0.05 plants /year (large capacity). Any larger construction rate could reduce the spent fuel inventory destined for storage. These and additional results will be presented.
Near Term Deployment, Long Term Impact: Uranium Price Over the Lifetime of New Capacity
Authors: Schneider Erich A. , Shah Neil
Published: July 2009

While reasonable short-term resource price projections can be obtained by taking a bottom-up approach — constructing a supply curve based upon current production capacities and costs — this approach breaks down as the time horizon of the analysis lengthens. One approach to long-term price forecasting is to calibrate a simple model of a commodity market against past data. To that end, an analogy was drawn between the behavior of the uranium market and that of some three dozen materials for which the United States Geologic Survey (USGS) maintains data. This work adds to previously published results showing that the USGS-reported prices of minerals similar to uranium have consistently declined over the past century. In this paper, the extent to which uranium geology and extraction technologies are indeed analogous to other minerals is quantitatively addressed. A study of crustal abundances, ore grades being economically mined, concentration factors, market share of extraction techniques, years of proven reserve and other factors indicates that uranium is not at all exceptional with respect to the average of the USGS minerals. This suggests that, on the supply side, the analogy between the USGS minerals and uranium may indeed offer valuable insights into medium and long term uranium price behavior.
Hot Rod Statistical Method for Large Break Loss of Coolant Accident With CATHARE
Authors: Keldenich Magali , Barbier Anthony , Grouhel Marie-Christine , Pignoly Ce´line
Published: July 2009

For nearly 30 years, Areva NP has contributed, with CEA, EDF and IRSN, to the development of CATHARE, one of the most advanced computer codes in the field of thermal-hydraulics calculations on the system-wide scale, with the objective to have at its disposal a high-performance tool, as the basis for the development of best-estimate methods for design and safety studies. At the same time, Areva NP has conceived with EDF a generic safety analysis method called Deterministic Realistic Method (DRM), which has been applied to elaborate and license ECCS evaluation models on the basis of the current CATHARE 2 V1.3L code version. DRM has then been extended by Areva NP with the Hot Rod Statistical Method (HRSM) which adds a statistical treatment of the hot rod-related uncertainties to the deterministic reactor system thermal hydraulic calculation and that can be used with other LB LOCA evaluation models. With the HRSM, the value of the peak cladding temperature envelope with a 95% probability level is determined based on direct Monte Carlo calculations statistically combining three key hot rod parameters randomly varied according to their probability distribution. The final value of the peak cladding temperature is then calculated taking into account an uncertainty calculated using the Bootstrap method. The HRSM is a method, based on CATHARE that is easy to implement as a complement to LB LOCA evaluation model. The large number of direct calculations performed ensures a very high degree of precision when evaluating the peak cladding temperature and the equivalent cladding reacted. The HRSM is currently used and licensed in France for the LB LOCA transients, and is now extended to SB LOCA transients using CATHARE 2 v2.5. An illustrative comparison of HRSM versus DRM typical results for Areva NP 3-loop and 4-loop plants then underlines the significant margins provided by this method.
Experimental Investigations on the Isolation Condenser Performance in Integral Test Loop
Authors: Sharma Manish , Pilkhwal D. S. , Vijayan P. K. , Saha D.
Published: July 2009

The proposed Advanced Heavy Water Reactor (AHWR) is a light water cooled and heavy water moderated pressure tube type boiling water reactor based on natural circulation. AHWR adopts several passive concepts with a view to simplify the design and to enhance safety and public acceptability. One such feature is passive decay heat removal using isolation condenser (IC) system during a station blackout. A scaled Integral Test Loop (ITL) was set up in BARC to simulate the overall system behavior studies for Advanced Heavy Water Reactor (AHWR). This facility simulates the Main Heat Transport System (MHTS), Emergency Core Cooling System (ECCS) and Isolation Condenser system (ICS) system, Feed Water System (FWS) and the associated controls. Power to volume scaling philosophy has been adopted for the design of the ITL systems. To evaluate the performance of IC, experiments have been carried out in ITL. The test results have been simulated using RELAP5/ MOD3.2. This paper deals with the experiments conducted, nodalization scheme adopted for ITL in RELAP5/MOD 3.2 simulation, transient predictions made and the results obtained in detail.
Benchmark Calculations of Pressurizer Model for Maanshan Nuclear Power Plant Using TRACE Code
Authors: Cheng Yi-Hsiang , Shih Chunkuan , Wang Jong-Rong , Lin Hao-Tzu
Published: July 2009

Pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) power plants. An accurate modelling of the pressurizer is needed to determine the pressure histories of the primary coolant system, and thus to successfully simulate overall PWR power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: 1) turbine trip test from 100% power; 2) large-load reduction at 100% power; 3) net-load trip at 100% power; and 4) net-load trip at 50% power. The simulation results are in reasonable agreement with the start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.
The Method of Characteristics for Arbitrary Geometry Based on the Customization of Autocad
Authors: Chen Qichang , Wu Hongchun , Cao Liangzhi
Published: July 2009

The method of characteristics (MOC) solves the neutron transport equation along each characteristics line. It is independent of the geometry shape of boundary and region. So theoretically, this method can be used to solve neutron transport equation in highly complex geometry. However, we should describe the geometry strictly and calculate intersect points of each characteristics line with every boundary and region firstly. In complex geometry, due to the great difficulty of treating of the arbitrary domain, the selection of geometry shape and efficiency of ray tracing are limited. So in most of the existing MOC codes, the type of outer boundary is limited as rectangular, hexagonal, triangular or the combination of them. And in order to treat reflective boundary condition, the selection of the azimuthal angles number is not freewill for different outer boundary shapes. To wipe off these limitations for using MOC in more complex fuel assembly calculation, the more powerful geometry treatment method is needed. The geometry treatment through the customization of AutoCAD, famous computer aided design software, is given in this paper. Thanks to the powerful capability of AutoCAD in graphics, the description of complex geometry becomes quite convenient. Moreover, with the language VBA (Visual Basic for Applications), AutoCAD can be customized to calculate intersect points of each characteristics line with every boundary and region. Thereby, no matter how complex or what kind of the geometry shape is, it can be treated uniformly by this method as long as it can be drawn in the AutoCAD. In addition, a directional interpolation method for the treatment of reflective boundary condition is introduced in this paper, so the number of azimuthal angles can be selected freely.
Transmission Probability Method Based on Triangular-Z Mesh for Solving Neutron Transport Equation in Three-Dimensional Geometry
Authors: Liu Guoming , Wu Hongchun
Published: July 2009

This paper presents a transmission probability method (TPM) to solve the neutron transport equation in three-dimensional triangular-z geometry. The neutron source within the mesh is assumed to be spatially uniform and isotropic. On the mesh surface, the constant and the simplified P1 approximation are invoked for the anisotropic angular neutron flux distribution. Based on this model, a code TPMTDT is encoded. It was verified by three 3D Takeda benchmark problems, in which the first two problems are in XYZ geometry and the last one is in hexagonal-z geometry. The numerical results of the present method agree well with those of Monte-Carlo calculation method and Spherical Harmonics (PN) method.
Modeling of Hydrogen-Air Detonations in the TONUS CFD Code and Its Application to the FLAME F-19 Test
Authors: Takasuo Eveliina
Published: July 2009

In severe accident management, the ability to predict pressure and thermal loads resulting from hydrogen combustion is important since they may threaten containment integrity. In computational modeling, different combustion regimes have to be accounted for and state-of-the-art techniques developed for reliable analysis. In the present study, the focus is on computational fluid dynamics code validation for reactive flows in the detonation regime. The FLAME hydrogen combustion test F-19 performed at the Sandia National Laboratories has been simulated by using the gas detonation model implemented in the TONUS CFD code which is developed by CEA and IRSN (France). In this model the reactive Euler equations are solved and the reaction rate is obtained by the Arrhenius global rate equation. Several simulations were run in order to examine the effect of modifying the parameters of the chemistry model. A mesh convergence study was performed for the purpose of finding out the necessary mesh resolution which could capture the detonation propagation with adequate accuracy. In addition, Chapman-Jouguet post-shock equilibrium conditions and the ZND detonation structure for the present gas mixture were examined by chemical kinetics calculations. The CFD simulation results were compared to the test results and the Chapman-Jouguet post-shock conditions. It was observed that the computational results differ from the C-J results with the C-J velocity being slightly exceeded. The model parameter study showed that it is not possible to significantly affect the flame propagation by adjusting the model parameters.
Potential Benefits of the New Westinghouse LOCA Mass and Energy Release Methodology: Estimated Reduction in the Initial Ice Mass Required for an Ice Condenser Containment
Authors: Ofstun Richard P. , Antolic Thomas F.
Published: July 2009

Westinghouse has developed a new loss of coolant accident (LOCA) mass and energy (M&E) release methodology using the WCOBRA/TRAC (WC/T) and GOTHIC computer codes. The new method replaces the deterministic post-blowdown steam generator and metal energy release models with a mechanistic calculation that results in significantly lower steam generator and metal energy release rates. As shown in Reference 1, these lower energy release rates result in a significantly lower post-blowdown containment pressure and temperature for the large dry containment design. This paper estimates the impact of the lower calculated post-blowdown energy release rates on the initial ice mass required to be loaded into an ice condenser containment.
Mode Analysis of Coolant Pump in Nuclear Power Plant
Authors: Ren Zhaohui , Ma Hui , Wang Xuejun , Wen Feng , Wen Bang Chun
Published: July 2009

With the reactor coolant in the nuclear power plant taken as the subject for research. A model of finite element is set up after the structure of the nuclear pump has been reasonably simplified. The natural frequencies of the pump with and without connecting rod are calculated so that the natural frequency and mode shape are explained briefly. The influence of snubber, tie rod and column on the natural frequency of the nuclear pump is computed. The analytic results can be used as the theoretical base for further earthquake response analysis of the nuclear pump.
Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal
Authors: Jeong Hae-Yong , Ha Kwi-Seok , Chang Won-Pyo , Lee Yong-Bum , Hahn Dohee , Cho Chungho
Published: July 2009

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.
Development of Simulation Codes for Some Accidents in Pressurized Water Reactors
Authors: Yang X. B. , Su G. H. , Qiu S. Z.
Published: July 2009

An analysis code has been developed for evaluating the transient thermo-hydraulic behaviors of the pressurized water reactor system. A series of mathematical and physical models is considered in this code, such as the point reactor neutron kinetics model, the heat transfer model, the friction model, the thermo-physical property model and so on. All possible flow and heat transfer conditions in some accidents have been considered and their corresponding models are supplied. Gear’s method is adopted for a better solution to the stiff equations. In this paper, some general accidents in the pressurized water reactors have been investigated, including the station blackout accident (SBO), the loss of flow accident (LOFA), the loss of feed water accident (LOFWA) and the reactivity insertion accident (RIA). The calculated results have been verified by the RELAP5/Mod3 and the results are satisfactory.
A Study on Bursting Pressure of Thin-Walled Cylinders
Authors: Deng Yang-Chun , Chen Gang , Sun Liang , Xu Tong
Published: July 2009

In this paper, we investigated the bursting pressure of thin-walled cylinders. Considering the strain hardening behavior of materials and the geometry deformation of pressure vessels, we derived the instability strain of thin-walled cylinders with a Swift-type stress-strain relationship, and used it as a failure criterion. Consequently, the instability stress was obtained and used to determine the maximum load-bearing capacity of thin-walled cylinders, that is, bursting pressure. The analytical solutions were compared with finite element analysis and bursting experimental results on different size thin-walled cylindrical pressure vessels manufactured from three different materials. It was turned out that it is reasonable to adopt instability strain as a failure criterion and use instability pressure as burst pressure. In the finite element analysis, the material parameters used were from raw experimental data or fitted values of experimental data. For both cases, finite element predications on instability strain and bursting pressure gave around the same values, close to experimental results. Therefore, based on finite element analyses, the instability strain and bursting pressure can be calculated by using true stress-strain curves directly measured from experiments, without the need to assume any specific material type.
New Melter Technology Development in Tokai Vitrification Facility
Authors: Aoshima Atsushi , Ueno Tsutomu , Shiotsuki Masao
Published: July 2009

Concentrated high radioactive liquid waste (HALW) containing almost FP and TRU elements is stored temporarily in Tokai Reprocessing Plant (TRP) and then transferred and vitrified in Tokai Vitrification Plant (TVF). TVF entered hot operation in 1995 about twenty years later of the start of hot operation in TRP. In TVF, total 247 canisters have been produced so far and JAEA should vitrify a large amount of HALW stored in TRP hereafter. Through about ten year operation of TVF, JAEA have experienced a failure of the first melter and got a lesson that stable operation of a melter and development of a melter with long life time were key points for high performance operation of TVF. Based on these experiences, JAEA has made a basic strategy to achieve stable TVF operation; keeping stable operation of the existing melter preventing adverse effect by noble metal accumulation and developing a new advanced melter with long lifetime preparing for future exchange as the third melter. According to the basic strategy, JAEA made a decade development plan of necessary key technologies two years ago [1] and has been continuing development. In this paper, present situation of the development is described.
Decay Heat Issues for Best Estimate Models
Authors: Sollima Calogero , Petrangeli Gianni
Published: July 2009

New capabilities of computational tools, recent demands from the industry, including nuclear power plant power uprating, and advancement in the understanding of safety concepts may require innovative solutions and approaches for safety analyses. In the framework of such activities, the purpose of the present paper is to streamline the results related to DH (Decay Heat) and the γ (gamma) – DH distribution, two of the relevant issues. Concerning the DH to be introducing in the calculations, a study is presented on the values of the possibility offered by present knowledge and RELAP 5 options; suggested also made on the assumptions to be used in the RELAP5 analysis. The second issue, that is dealt with in this paper, is the DH γ distribution. The short study presents the effect of considering the real absorbed γ distribution in the core. The importance of taking into account this effect is apparent when the influence on important calculation results is considered: e.g. PCT (Peak Clad Temperature) could be affected by more than 100 K.
Development of Remote Measuring System for RVI Modularization
Authors: Kim Tae-Soon , Lee Jae-Gon , Lee Je-Jun , Park Myeong-Man
Published: July 2009

The construction duration of a nuclear power plant has been considered as a important factor to occupy the competitive edge. For the optimization process of APR1400 which is nuclear reactor newly developed in Korea, it has been suggested that the modularization of reactor vessel internals (RVI) was one of useful means to reduce the construction duration. In general, RVI consists of three components such as core support barrel (CSB), lower support structure/core shroud (LSS/CS) and upper guide structure (UGS). It is complicated and tedious to assemble the RVI by the conventional method which requires about 8∼10 months. In order to modularize the RVI, the gap between the CSB snubber lug and the reactor vessel (RV) stabilization lug must be measured by a remote measurement method. By using a remote measurement method, the welding of CSB and LSS/CS can be performed in advance of the reactor installation process to reduce the construction duration of a nuclear power plant. Compared with the conventional method, the duration of about 2 months required in the welding of CSB and LSS/CS is finally reduced. In this study, first of all we developed the remote measuring system that included the digital probes to measure the 72 points of gap at once. The system device consists of digital probe section, pneumatic supply and control section, electric power section, remote control computer and program. The selected digital probe of linear variable differential transformer (LVDT) type and the calibration device for the zero-point adjustment jig and the other devices have sufficient reliability and accuracy. And the digital probe connection jig has sufficient consistency. The network and system for remote measurement were very stable and no disturbance at electromagnetic interference environment. And we carried out the proof test of our remote measuring system to evaluate the application on the real plant conditions using the RV and RVI mock-up. The results of remote measurement were compared with existing manual measuring method and the reliability of the system was verified. Finally, we confirmed that our remote measuring system had the efficient reliability could be applied to measure the gap of RVI.