bluuebunny/crossref_metadata_2025 huggingface dataset:
Modelling of Fe-Cr Martensitic Steels Corrosion in Liquid Lead Alloys
Authors: Balbaud-Ce´le´rier F. , Martinelli L.
Published: February 2010

Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience. However, the availability and the acceptability of the SFR are affected by the problems linked with the sodium-water reaction. One innovative solution to this problem is the replacement of the sodium in the secondary loops by an alternative liquid fluid. Among the fluids considered, lead-bismuth is at the moment being evaluated, liquid lead-bismuth has been considerably studied in the frame of the research program on Accelerator Driven Systems for transmutation applications. However, lead alloys are corrosive towards structural materials. The main parameters impacting the corrosion rate of Fe-Cr martensitic steels (considered as structural materials) are the nature of the steel (material side), the temperature, the liquid alloy velocity and the dissolved oxygen concentration (liquid alloy side). In this study, attention is focused on the behaviour of Fe-9Cr steels and more particularly T91 martensitic steel. It has been shown that in the case of Fe-Cr martensitic steels the corrosion process depends on the concentration of oxygen dissolved in Pb-Bi. - For an oxygen concentration lower than the one necessary for magnetite formation (approximately < 10−8 wt% at T ≈ 500 °C for Fe-9Cr steels), corrosion proceeds by dissolution of the steel. - For a higher oxygen content dissolved in Pb-Bi, corrosion proceeds by oxidation of the steel. These two corrosion processes have been experimentally and theoretically studied in CEA Saclay and also by other partners leading to some corrosion modelling in order to predict the life duration of these materials as well as their limits of utilisation. This study takes into account the two kinds of corrosion processes, dissolution and oxidation. In these two different processes, the lead alloy physico-chemical parameters are considered: the temperature and the liquid alloy velocity for both processes and the oxygen concentration for oxidation.
Verification of Discontinuous Random Walk Lagrangian Particle Tracking as a Tool to Model Deposition in Annular Flow
Authors: Caraghiaur Diana , Anglart Henryk
Published: February 2010

A verification of the applicability of the discontinuous random walk (DWR) Lagrangian Particle Tracking (LPT) model to calculate deposition in annular two-phase flow has been conducted. The comparison of simulation results to experimental data of deposition of mono-sized droplets shows that the model follows the correct trend in inertia-moderated regime, but is un-reliable in diffusion-impaction deposition regime. The comparison to other experimental studies of annular flow of different density ratios between the two phases reveals that the density ratio is incorrectly incorporated into the model, since the experimental trend is reversed. It can be concluded that the applicability of DWR LPT model for deposition calculation in steam-water flow at BWR conditions cannot be validated by solely comparing the simulation results to air-water experimental data.
Numerical Prediction of Void Distribution in Two-Phase Turbulent Flow in a Sub-Channel
Authors: Ikeno Tsutomu , Sasakawa Tatsuya , Kataoka Isao
Published: February 2010

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.
Experimental and Numerical Investigations of Temperature Stratification Phenomena in Passive Safety Systems for Decay Heat Removal
Authors: Krepper Eckhard , Beyer Matthias
Published: February 2010

Modern concepts of nuclear power reactor systems are equipped with passive systems for decay heat removal. Examples are the pool of the emergency condenser (BWR-1000) or the pool of the ESBWR. These systems operate without active influence from outside. The questions arise: How reliable are the based physical mechanisms? Are they understood completely? Are actual models able to describe the phenomena? In different passive systems the energy is transferred by natural circulation into large pools which are considered as infinite heat sink. The paper deals with experiments and with CFD simulations to investigate the capability of actual CFD codes to describe these phenomena. In the FZ Dresden-Rossendorf at the facility TOPFLOW heating-up tests of an emergency condenser were performed. During these tests also the temperature courses on the secondary side of the pool were recorded. The data recording comprises periods starting from single phase liquid until steam on the secondary pool side was found. During these experiments temperature stratification phenomena were observed, which were found in earlier small scale tests. In the paper also these small scale experiments are described. A detailed CFD analysis of these experiments was performed. An explanation of the observed phenomena on the basis of the small scale tests and the CFD simulations is presented.
Experimental Study on Influence of Interfacial Behavior on Jet Surface Fragmentation
Authors: Uchiyama Yuta , Abe Yutaka , Kaneko Akiko , Nariai Hideki , Yamagishi Makoto , Matsuo Eiji , Koyama Kazuya , Itoh Kazuhiro
Published: February 2010

For the safety design of the Fast Breeder Reactor (FBR), it is strongly required that the Post Accident Heat Removal (PAHR) is achieved after a hypothetical Core Disruptive Accident (CDA). In the PAHR, it is important that the molten material is fragmented to be solidified by the sodium coolant with high boiling point and thermal conductivity. Furthermore, in order to estimate whether the molten material jet is completely solidified in sodium coolant or not, it is necessary to evaluate the jet breakup length. Although there are many previous studies on the jet breakup length, the tendency of jet breakup length is different for the previous studies. To estimate jet breakup length, it is necessary to understand the interaction between molten core material and coolant. The objective of the present study is to clarify the influence of the interfacial behavior of the jet on the fragmentation behavior on the jet surface. The experiments are conducted to obtain the interfacial behavior and the fragmentation behavior on the jet surface by injecting transparent Fluorinert™ (FC-3283) into water. The jet breakup behavior of the Fluorinert and the fragmentation behavior on the jet surface in pool are observed by using high speed video camera. To clarify the influence of interfacial behavior on jet surface fragmentation, it is necessary to clarify the effect of the internal flow of the jet and the surrounding flow structure on the interfacial behavior. The internal and the external velocity distribution of the jet are obtained by Particle Image Velocimetry (PIV) technique from the visual data. Shear stress is evaluated from the velocity data obtained by PIV technique. Reynolds stress and turbulent energy are also evaluated from the velocity data. As the results, shear stress becomes large along the interfacial wave. The maximum value of shear stress is decreased toward downstream. Reynolds stress becomes large at the jet surface. The vortex around the interfacial wave is observed by PIV measurement. The local shear stress acts on the interfacial wave. It is suggested that the local shear stress on the jet surface causes the fragmentation. From the experimental results, the interaction between the interfacial behavior of the jet and flow structure of the jet and surrounding fluid are discussed. The dominant mechanism of the fragmentation behavior and the influence of local shear stress at the interface on the fragementation are also discussed.
Hydrodynamic Modeling of Mineral Wool Fiber Suspensions in a Two-Dimensional Flow
Authors: Cartland Glover Gregory M. , Grahn Alexander , Krepper Eckhard , Weiss Frank-Peter , Alt So¨ren , Hampel Rainer , Ka¨stner Wolfgang , Kratzsch Alexander , Zacharias Frank
Published: February 2010

A consequence of a loss of coolant accident is that the local insulation material is damaged and maybe transported to the containment sump where it can penetrate and/or block the sump strainers. An experimental and theoretical study, which examines the transport of mineral wool fibers via single and multi-effect experiments is being performed. This paper focuses on the experiments and simulations performed for validation of numerical models of sedimentation and resuspension of mineral wool fiber agglomerates in a racetrack type channel. Three velocity conditions are used to test the response of two dispersed phase fiber agglomerates to two drag correlations and to two turbulent dispersion coefficients. The Eulerian multiphase flow model is applied with either one or two dispersed phases.
More on Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC
Authors: Zeng Lingfu , Jansson Lennart G. , Dahlstro¨m Lars
Published: February 2010

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. Our work presented earlier in ICONE16 categorized the design rules in ASME III into linear design and non-linear design rules and examined the corresponding design requirements. In this paper, a more in-depth review of these rules for service limit level D is conducted. In particular, several rules which are currently intensively applied but often found to be inconsistent and confusing in Sweden are studied and discussed in detail. Suggestions for improvements and guides for a reasonable application of these rules, which have been practiced in several ongoing projects, are given.
CFD Study on Mixing in VVER-440 Fuel Rod Bundle With Guiding Vanes in Spacer Grids
Authors: Brandt Tellervo , Toppila Timo
Published: February 2010

In this article, we study mixing in a fuel rod bundle geometry used in VVER-440 type pressurized water reactors. The flow inside the fuel assembly under normal operation is simulated using the FLUENT 6.3 computational fluid dynamics (CFD) solver. Steady state one phase Reynolds-averaged Navier-Stokes modeling is applied with two-equation turbulence models. In the first part of the article, grid independence is studied using a small submodel of the fuel rod bundle. In the second part, flow across a full-length fuel rod bundle is simulated both with constant temperature and with prescribed heat fluxes at the rod walls. Finally, in the third part of the article, we study how guiding vanes included in the spacer grids would enhance mixing in the hottest subchannels. We propose that having four guiding vanes on each side of the hexagonal spacer grid would produce most mixing in the desired locations of the fuel rod bundle.
Making FAC Calculations With BRT-CICERO™ and Updating to Version 3.0
Authors: Trevin Ste´phane , Persoz Matthieu , Knook Thomas
Published: February 2010

The surveillance of Flow Accelerated Corrosion (FAC) on secondary pipes is a major concern for every nuclear power plant operator. After the Surry accident in 1986, EDF launched a computer code development program to monitor this degradation phenomenon. A chemical corrosion model has been developed, based on laboratory test results obtained by EDF R&D since the late 70’s. This model enables to compute the wall thickness loss of pipes submitted to FAC, with respect to the thermo-hydraulic conditions, the fluid chemistry, the material chromium content, the pipe geometry and the cycles duration.
Recent Hydrodynamics Improvements to the RELAP5-3D Code
Authors: Riemke Richard A. , Davis Cliff B. , Schultz Richard R.
Published: February 2010

The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radionuclide transport model, (5) improved pump model, and (6) compressor model. These changes will be discussed.
Role of Deterministic Design Safety Criteria in Assuring Safety of Generation IV Reactors
Authors: Ahn Sang Kyu , Kim Inn Seock , Lee Hun-Joo , Hong Soon Joon
Published: February 2010

In the nuclear power community, deterministic design safety criteria have been used as a major means for assuring safety of nuclear power plants, e.g., light water reactors (LWRs). However, as a result of considerable advances in the quantitative risk analysis technique, such as Probabilistic Risk Assessment (PRA), risk-informed approaches are increasingly applied together with some of the deterministic approaches that are still considered valid. In this paper, the various deterministic approaches that have played an important role in ensuring nuclear power safety are critically reviewed to shed light to the necessary characteristics of the desirable design safety criteria especially with regard to advanced reactors such as Generation IV reactors that have great potential to further enhance the economic and safe production of nuclear power.
Investigation of the Agglomeration and the Break-Up of Isolation Material
Authors: Renger Stefan
Published: February 2010

The description of the complex flow of coolant water with particles is necessary to evaluate safety relevant effects of the sedimentation of isolation material on sump-sieves in nuclear power station. Classifying and modeling of the different phenomena maybe important in the case of a coolant accident, because the isolation material can be transported into the reactor containment, the building sump of the containment and into the associated systems [1]. In order to ensure the heat dissipation from the reactor core and the containment the cooling systems transport the water from the sump into the condensation chamber and then into the reactor pressure vessel. The functionality of the pumps can be affected by a high allocation of the sieves with fractionated isolation material. In this case the heat dissipation from fuel elements is not guaranteed. The transport of the material will be simulated with the CFD-code. The modeling of the flow with particles is very complex, because of the structure of the particles and their interaction with the fluid. There are different classes of particles with different attributes, e.g. sinking velocity. So one needs more than one disperse phase to describe the whole process, which is associated with a lot of computing power and not realizable for large geometries. The paper deals with experimental and methodical activities for the description of the agglomeration and the break-up of isolation material in fluid flow. The aim of this work is to describe the evolution of the volume parts of the different particle classes turbulent flows depending on the time in. The modeling phase starts with a very simple model to describe 3 particle classes (x, y, z) and results in a differential equation system with 3 equations. To describe all classes the model has to be expanded. Therefore the Lindenmayer-System approach has been adopted. These systems can be taken in cases where self-similarity takes place. The result is a differential equation system with iterations for the three classes (x(i), y(i), z(i)), with i as the parameter for the number of subclasses. The values for the agglomeration and break-up rates will be taken from experiments. As a result a model has been created which describes the evaluation of the different particles classes in turbulent flow. It helps to choose the correct particle class in the CFD simulation depending on the situation to simulate.
RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR
Authors: Li Chen-Lin , Tsai Chiung-Wen , Shih Chunkuan , Wang Jong-Rong , Chung Su-Chin
Published: February 2010

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.
Effects on Core Damage Frequency to Sensitivity of Instrumentation and Control Reliability
Authors: Kitzmiller John T. , Anderson Richard G. , Genutis Laura L. , Popp Dennis M.
Published: February 2010

Core Damage Frequency (CDF) and Large Release Frequency (LRF) can be greatly affected by the application of Instrumentation and Control (I&C) within the control and safety systems utilized in the operations of a nuclear power plant. It is important that the modeling of these systems be an accurate representation of the I&C design to capture the effects of the I&C on the operations of the nuclear power plant as far as control and safety that are needed for the effective and safe operation of power generation. Various sensitivities can be performed on I&C modeling in the AP1000 Probabilistic Risk Assessment (PRA) to determine the impact on CDF. These sensitivities can be in the areas of base reliability values (probabilities), common cause failures (CCF) of software as well as common cause (CC) factors of computer modules. Insights can be found as to the effects of these CCFs by varying the modeled probabilities in the most conservative direction to determine how the model will respond. The model response, i.e., CDF and cutset importance orders will provide insights as to how sensitive I&C modeling is to CCF and how sensitive the rest of the PRA model is to I&C. Increasing failure probability of the Protection and Safety Monitoring System (PMS) and Plant Control System (PLS), and Diverse Actuation System (DAS) to explore the minimum reliability that would support favorable CDF and LRF values, as well as, the effect of total failure of the PMS (no credit taken for PMS in core damage sequences) and PLS (no credit taken for PLS in core damage sequences) on CDF values are explored. Sensitivity analyses show that the CDF increases if no credit is taken for operator actions. For AP1000, this sensitivity study indicated a decrease in dependence on operator actions over conventional nuclear plants. This conclusion is likely a result of the increased reliability of the PMS to automatically actuate the given systems and components. For this reason, the most important system is PMS, in the case where no credit is taken for PMS in core damage sequences.
AP1000 Plant Availability Analysis in Accordance With the EUR
Authors: Anderson Richard G. , Conner Shannon L. , Demetri Kathryn J. , Holla¨nder So¨nke , Maioli Andrea , Timmons Thomas F.
Published: February 2010

As part of the European Passive Pressurized Water reactor (EPP) Program, AP1000 plant availability analyses and studies were performed to show compliance with the European Utility Requirements (EUR) goal of an overall plant availability greater than 90%. The EUR annual Design Availability Factor averaged over a 20-year period is defined as a function of the reference duration and number of a pre-defined set of normal outages (e.g., regular refueling and maintenance, 10 years in-service inspection, turbine overhaul) as well as a function of expected forced outages. Reference duration for the regular outages has been evaluated on a deterministic basis using current experience tailored for the AP1000 18 month fuel cycle. Provision for special work has been addressed, also in a deterministic fashion, by assuming a steam generator replacement. Estimation of the annual forced outage has been addressed in a two-fold probabilistic fashion: the expected unavailability due to generic balance of plant and support systems have been derived on a system-level basis from historic operational data, relying on AP600 based information and on updated European operating plant historical information provided by European utilities participating in the EPP Program. Such system-specific analyses also addressed the evolution from the AP600 to the AP1000 design as well as EPP-specific issues (such as the increased Reactor Coolant Pumps (RCP) unavailability due to continuous use of a variable speed driver on a 50 Hz grid). A selected set of AP1000 front end systems have then been analyzed with a fully probabilistic approach by means of Fault Tree Analysis (FTA) and Reliability Block Diagram (RBD) techniques, using reliability data (i.e., failure rates and Mean Time To Repair (MTTR)) for basic components. Again, the evolution of the design from the AP600 and the previous EPP availability analysis has been factored into these studies. The result of the evaluation leads to an overall AP1000 Annual Design Availability of 93.7% averaged over a 20-year period. This exceeds the 90% EUR availability goal.
Evaluation of the Flow and Heat Transfer Characteristic in Vertical Narrow Annulus Channel Based on the RELAPSCDAPSIM/MOD3.4 Code
Authors: Liu Jiange , Peng Minjun
Published: February 2010

The physical phenomenon of the flow and heat transfer characteristic in vertical narrow annulus channel is a little different from general channel. It is valuable to improve the model veracity for exactly predicting the nuclear power system key physical phenomenon. The heat transfer characteristic of two narrow annulus channels (narrow gap 1.0mm and 1.5mm) is analyzed by means of the RELAPSCDAPSIM/MOD3.4 code recently developed by SDTP. Compared with related experiment data, the results show that the new developed code could predict key parameters such as fluid temperature, wall temperature, heat transfer coefficient, pressure drop, the occurrence condition of the flow instability and so on. The code has the ability to simulate the narrow annulus channel heat transfer characteristic.
Thermo-Hydraulic Calculation and Safety Analysis of the Tehran Research Reactor
Authors: Jafari J. , Firuzjaee M. K. , Khakshournia S. , Sepanloo K. , D’Auria F.
Published: February 2010

The aim of this study is to evaluate the thermal hydraulic parameters such as, Minimum DNBR, fuel, clad and coolant temperatures of the Tehran Research Reactor (TRR) core in order to assure that the thermal limitations are not exceeded at full power and steady-state and the loss-of-flow accident (LOFA) operating conditions and the existing design is adequate to assure that the consequences from this anticipated occurrence does not lead to a severe accident. The PARET and TRANS computer codes were used to analysis thermal-hydraulic and safety parameters under study steady-state conditions. Furthermore the LOFA as a consequence of two type of transients are analyzed; one related to main cooling pump failure accident (slow LOFA), and the other related to accidentally opening the flapper valve whereas the main cooling pump is in working (fast LOFA). The steady-state analysis involves the determination of the departure from nucleate boiling ratio (DNBR), the temperatures profile, heat flux across the hottest channel and etc. The LOFA analysis involves the power and coolant flow rate, the fuel centerline, clad surface and coolant outlet temperatures, and the occurring of coolant flow reversal.
Knowledge-Based Information Resource Management System for Materials of Fast Reactors
Authors: Hsieh Hsuan-Tsung Sean , Li Ning , Chen Yitung , Kwan Kenny , Huang Jen-Yuan , Lee Changyeol
Published: February 2010

In the development of advanced fast reactors, materials and coolant/material interactions pose a critical barrier for higher temperature and longer core life designs. For advanced burner reactors (sodium cooled), experience has shown that the qualified structural materials and fuel cladding severely limits the economic performance. In other liquid metal cooled reactor concepts, advanced materials and better understanding and control of coolant and materials interactions are necessary for realizing the potentials. Researches from universities, national laboratories and related industrial participants have been continuously generating invaluable data and knowledge about materials and their interactions with coolants in the past few decades. Under the consideration of cost and time constraints, the paradigm of designing and implementing a successful Gen IV Nuclear Energy Systems can be shifted and updated via the integration of information and internet technologies. Such efforts can be better visualized by implementing collective (centralized or distributed) data storages to serve the community with organized material data sets. Material property data provided by MatWeb.com and the ongoing development of web-based GEN IV material handbook are few examples. From system design perspective, sodium-cooled fast reactor (SFR) proposed in the GEN IV system have been significantly developed. According to the GEN IV ten-year program plan, current R&D work will be pointed to demonstration of the design and safety characteristics, and design optimization. All of those activities follow the path of data generation, analysis, knowledge discovery and finally decision making and implementation. We are proposing to create a modularized web-based information system with models to systematically catalog existing data and guide the new development and testing to acquire new data. Technically speaking, information retrieval and knowledge discovery tools will be implemented for researchers with both information lookup options from material database and technology/development gap analysis from intelligent agent and reporting components. The goal of the system is not only to provide another database, but also to create a sharable and expandable platform-free, location-free online system for research institutes and industrial partners.
Optimization and Analysis for Passive Residual Heat Removal System of an Integral Pressurized Water Reactor
Authors: Dai Shoubao , Peng Minjun , Liu Jiange
Published: February 2010

The characteristics of passive safety systems for an integral pressurized water reactor (IPWR) are quite different from the general reactor because of special configuration and dangerous run environment. Passive residual heat removal system (PRHRS) for the IPWR with three-interknited natural coolant circulation loops, safely remove the core decay heat to the ultimate heat sink. Using RELAP5/MOD3.4 code to simulate this system during the reactor trip, analyses the steady-state and transient-state thermohydraulic behaviors for the IPWR and its PRHRS, and the effects of design parameters on the system. it is found that on the initial period of reactor trip, due to the establishment of the natural circulation in three loops uncompleted, the secondary loop pressure have the peak value. Through analyzing the effects of design parameters on the system, the PRHRS are optimized. The results show that the larger the residual heat exchanger (RHE) heat transfer area and the higher the height difference between the steam generator and the residual heat exchanger, the easier the establishment of the natural circulation in the third loop, but which make the peak value of the secondary loop pressure higher. According to set the compensating water tank, which is parallel connected to the RHE, can lighten the higher the peak value of the secondary loop pressure, and optimize the design of PRHRS.
Review and Impact Evaluation of ASME NQA-1 (2008)
Authors: Wilde Taunia , McEvoy Tim , Holmes Richard , Sandquist Gary M.
Published: February 2010

ASME has issued a 2008 revision to the Nuclear Quality Assurance Standard, NQA-1 that impacts the siting design, construction, startup and operations of new generation nuclear power plant designs as well as other nuclear facilities. In view of new nuclear plants recently certified by the NRC, the NQA-1 2008 revision is poised to meet those QA issues and requirements that have or may arise during ESP, COL and other regulatory actions by the NRC. In view of the enhanced safety features and significant design changes associated with this new generation of nuclear infrastructure including the DOE development of the CMRR (Chemical and Metallurgy Research Replacement) at Los Alamos, present QA programs and procedures require a re-evaluation and assessment if the 2008 revision of NQA-1 is widely adapted in the US and possibly other countries. A synopsis of the revisions posed by the 2008 revision to former QA standards is given together with ancillary impacts for the nuclear community.
Sensitivity Analysis of Kinetics Parameters of Tehran Research Reactor (TRR)
Authors: Hosseini Seyed Abolfazl , Vosoughi Naser
Published: February 2010

In this research, effective delayed neutron fraction (βeff) and neutron generation time (Λ) of the Tehran Research Reactor (TRR) are calculated for different uranium enrichments from 14.84 w/o to 96.56 w/o U235 in two states of the TRR, (cold fuel, clad and coolant temperature of 20 °C; and hot fuel, clad and coolant temperature of 65, 49 and 44 °C, respectively) using the MTR_PC computer code. Comparative analysis shows that both βeff and Λ increase as fuel enrichment decreases. However, variation rate of βeff is not the same in two conditions. βeff in the hot state is larger than those calculated in the cold state when fuel enrichment goes more than 83.91%, while the situation is reverse for enrichment less than that. The obtained neutron generation time shows normal behavior for all different fuel enrichments. The variables involved in kinetics parameters calculations (i.e., neutron fission cross section, fuel enrichment, etc.) are investigated theoretically to confirm the results of calculations in cold and hot states. Variations of βeff and Λ with fuel burnup are studied too.
Status and Near-Term Objectives of the Works on the EUR Document
Authors: Berbey Pierre , Hedin Franc¸ois , VanHoenacker Luc
Published: February 2010

In 2007–2008, the European Utility Requirements (EUR) works have been focused on volume 3 (evaluation of the available Gen 3 designs) and volume 4 (conventional island). The works on the AP1000 and AES92 subsets of the EUR volume 3 have been concluded at the end of 2007. The texts have been published and are now available for the EUR members and other utilities. The works on the EPR subset of volume 3 have resumed in 2007. A revision B is being produced for which representatives from ten EUR utilities and from Areva NP have been involved in the revision of the analysis of compliance. Meetings of the specific EUR coordination group in charge of this task have been organized every 4–5 weeks throughout 2008. The revised version of the EPR subset of the EUR volume 3 should be finalized around mid 2009. The revision C of the EUR volume 4 is now available after a thorough review has been performed within the EUR organization to make it consistent with the revision C of the EUR volume 2 published in 2001. A lot of preparatory material for a possible revision D of the EUR volumes 1 and 2 has been produced since 2002. Since important contributions are not yet available the decision to integrate this revision D is still to come. The EUR organization has kept enlarging: Energoatom, ENEL and Endesa have been welcomed as full members; CEZ and MVM are now EUR associated members. New LWR projects of potential interest for the EUR utilities are being contemplated. For instance a preliminary assessment of compliance of MHI’s APWR project has been worked out in the first months of 2008. Recently the EUR and ENISS organizations have decided to join their efforts in a collaboration scheme in which they will coordinate their positions and actions in nuclear safety with respect to the LWR Gen3 designs. The two organizations will cooperate in their relations with the other stakeholders, in particular with the IAEA and WENRA organizations. In addition EUR and CORDEL (Cooperation in Reactor Design Evaluation and Licensing), which is a WNA (World Nuclear Association) working group decided also to coordinate their efforts for the industry benefit, in relation with the MDEP (Multinational Design Evaluation Program) initiative of safety nuclear regulators.
Next Generation Safety Analysis Methods for SFRs—(5) Structural Mechanics Models of COMPASS Code and Verification Analyses
Authors: Shirakawa Noriyuki , Uehara Yasushi , Naitoh Masanori , Okada Hidetoshi , Yamamoto Yuichi , Koshizuka Seiichi
Published: February 2010

A five-year research project started in FY2005 (Japanese Fiscal Year, hereafter) to develop a code based on the Moving Particle Semi-implicit (MPS) method for detailed analysis of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). CDAs have been almost exclusively analyzed with SIMMER-III [2], which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code, coupled with fuel pin model and neutronics model. The COMPASS has been developed to play a role complementary to SIMMER-III in temporal and spatial scale viewpoint; COMPASS for mesoscopic using a small window cut off from SIMMER-III for macroscopic. We presented the project’s outline and the verification analyses of elastic structural mechanics module of the COMPASS in ICONE16 [1]. The COMPASS solves physical phenomena in CDAs coupling fluid dynamics and structural dynamics with phase changes, that is vaporization/condensation and melting/ freezing. The phase changes are based on nonequilibrium heat transfer-limited model and all “phase change paths” considered in SIMMER-III are implemented [20]. In FY2007, the elastoplastic model including thermal expansion and fracture are formulated in terms of MPS method and implemented in the COMPASS, where the model adopts the von Mises type yield condition and the maximum principal stress as fracture condition. To cope with large computing time, “stiffness reduction approximation” was developed and successfully implemented in the COMPASS besides parallelization effort. Verification problems are set to be suitable for analyses of SCARABEE tests, EAGLE tests and hypothetical CDAs in real plants so that they are suggesting issues to be solved by improving the models and calculation algorithms. The main objective of SCARABEE-N in-pile tests was to study the consequences of a hypothetical total instantaneous blockage (TIB) at the entrance of a liquid-metal reactor subassembly at full power [21]. The main objectives of the EAGLE program consisting of in-pile tests using IGR (Impulse Graphite Reactor) and out-of-pile tests at NNC/RK are; 1) to demonstrate effectiveness of special design concepts to eliminate the re-criticality issue, and 2) to acquire basic information on early-phase relocation of molten-core materials toward cold regions surrounding the core, which would be applicable to various core design concepts [22, 23]. In this paper, the formulations and the results of functional verification of elastoplastic models in CDA conditions will be presented.
Foundation Engineering for Nuclear Plants: U.S. vs. European Code and Practice
Authors: Massoudi Nasser
Published: February 2010

This paper reviews the current regulatory and industry practices in geotechnical investigations for nuclear power plants in the U.S. and Europe, with the intent to highlight the common features and underscore the differences. Specifically, applicable sections of regulatory and industry-established codes and practices are reviewed as relate to geotechnical practices and foundation engineering. Similarly, regulatory requirements such as those established by the U.S. Nuclear Regulatory Commission and the European equivalents will be reviewed. The paper serves as a vehicle to highlight industry and regulatory common grounds, as well as variations in the two practices, in the spirit of disseminating knowledge on codes and standards and facilitating international cooperation between the foundation engineering community in the U.S. and Europe.
Next Generation Safety Analysis Methods for SFRs—(1) Brief Introduction of the Project and Basic Study for Algorithm of Particle Method
Authors: Koshizuka Seiichi
Published: February 2010

A five-year research project started in FY2005 to develop a code based on the Moving Particle Semi-implicit (MPS) method for detailed analysis of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). Separated phenomena in CDAs are focused on. Both mixed-oxide (MOX) and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are investigated by phase diagram calculation, classical and first-principles molecular dynamics. Basic studies concerning the numerical methods are performed to support the code development of COMPASS. In the validation process of COMPASS, calculations using SIMMER-III are carried out to give boundary conditions and to provide reference results. In this paper, the overall status of the project in FY2007 is provided. Two basic studies for the particle method are also described. One is the development of a unified algorithm for compressible and incompressible flow. The developed algorithm is tested by a benchmark problem of a shock tube. The result agrees well with the analytical solution. The other is the development of a turbulence model for the particle method. Large Eddy Simulation (LES) is selected. We developed a wall boundary condition to reproduce the log law near the wall with a coarse spatial resolution. Flow over a backward facing step is calculated by the present model. The result agrees with an experimental data.
Influence of System and Neutron-Kinetic Parameter Variations on an Anticipated Transient Without SCRAM in a PWR
Authors: Kliem So¨ren , Mittag Siegfried , Rohde Ulrich , Weiss Frank-Peter
Published: February 2010

The complete failure of the reactor scram system upon request during an operational transient is called anticipated transient without scram (ATWS). According to the German regulatory guidelines, postulated ATWS events have to be analyzed with regard to their consequences on the safety of nuclear power plants. Since the course of ATWS transients is determined by a strong interaction of the neutron kinetics with the thermal hydraulics of the system, coupled 3D neutron kinetic/thermal hydraulic code systems are adequate tools for the analysis of such transients. The coupled code system DYN3D/ATHLET developed at FZD is applied to the analysis of an ATWS transient. The objective of the present work is to perform a best-estimate analysis with consequent use of a 3D neutron kinetic code (DYN3D) in combination with an advanced thermal hydraulic system code (ATHLET) together with a quantification of differences in the course and the results of transients, which arise from the uncertainties of thermal hydraulic and neutron-physical conditions. Typically, the complete failure of the main feed water supply is assumed to be the bounding ATWS event with regard to the maximum primary coolant pressure, which can be reached during the transient. The limitation of the coolant pressure is a precondition for the integrity of the primary circuit. The situation is aggravated if the main coolant pumps remain in operation. For this particular transient, the influence of different thermal hydraulic and neutron-physical conditions on the course of the transient was analyzed. In a number of code runs all systems which have an influence on the course of the transient were varied. These are the auxiliary boration system and the auxiliary feed water supply. Further, the influence of the modeling of the pressurizer safety and relief valves as well as the steam bypass system on the secondary side was assessed. The variation of the pressurizer relief and safety valve behavior has the biggest influence on the primary circuit coolant pressure. In the second part, two different core loading patterns were generated for the analyses by varying the number of MOX (mixed oxide) fuel assemblies (FA) in the core. The basic core loading contains 64 MOX FA. All these MOX FA were replaced by standard uranium oxide FA. The presence of MOX in the core has a remarkable influence on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behavior of the coolant pressure in the first part of the transient. It has been demonstrated that the pressure maximum decreases with growing MOX portion in the core. The maximum pressure determined in the calculations with variation of system and neutron-physical boundary conditions does not reach the allowed limit for the primary circuit.
Next Generation Safety Analysis Methods for SFRs—(8) Analyses of Eutectics Between Fuel and Steel in Metal Fuel With FPMD Code VASP
Authors: Himi Masashi , Yamamoto Yuichi , Nagamine Yasuo , Shirakawa Noriyuki , Uehara Yasushi , Arima Tatsumi
Published: February 2010

There are two main objectives in this study. One is to estimate atomic diffusion coefficients in eutectic reaction between metal fuel and cladding materials in order to establish the atomic diffusion model for the COMPASS code. The other is to estimate their material properties such as Young’s modulus in high temperature up to near melting points in core disruptive accidents (CDAs) in Sodium-cooled Fast Reactors (SFRs). We used the first principle molecular dynamics (FPMD) code VASP to realize the two objectives. We tried to understand the initiation mechanism of eutectics based on change of electronic state energy accompanied by change of Kohn-Sham energy, including phonon effect. In this project [1], three methods, phase diagram calculation (CALPHAD), classical molecular dynamics (CMD), and FPMD, are employed to understand the mechanism of eutectics and to introduce dynamic characteristics in eutectic phenomena into the COMPASS code.
Research on Monte Carlo Perturbation Calculation Methods Applied in Reactor Physics
Authors: Li Ze-guang , Wang Kan , Yu Gang-lin
Published: February 2010

In the reactor design and analysis, there is often a need to calculate the effects caused by perturbations of temperature, components and even structure of reactors on reactivity. And in sensitivity studies, uncertainty analysis of target quantities and unclear data adjustment, perturbation calculations are also widely used. To meet the need of different types of reactors (complex, multidimensional systems), Monte Carlo perturbation methods have been developed. In this paper, several kinds of perturbation methods are investigated. Specially, differential operator sampling method and correlated tracking method are discussed in details. MCNP’s perturbation calculation capability is discussed by calculating certain problems, from which some conclusions are obtained on the capabilities of the differential operator sampling method used in the perturbation calculation model of MCNP. Also, a code using correlated tracking method has been developed to solve certain problems with cross-section changes, and the results generated by this code agree with the results generated by straightforward Monte Carlo techniques.
A Case History of CFD Support to Accelerator Driven System Plant Design
Authors: Moreau Vincent , Mansani Luigi , Petrazzini Maurizio
Published: February 2010

The Integrated Project EUROTRANS, funded by the European Commission in the VI European framework program, was aimed at providing the advanced design of a multi purpose research oriented Accelerator Driven System (ADS), called eXperimenTal-ADS (XT-ADS), and the preliminary design of an industrial scale ADS, called European Facility for Industrial Transmutation (EFIT). One contribution of CRS4 (Centro di Ricerca, Sviluppo e Studi Superiori in Sardegna) has been to provide support to the overall plant design by means of Computational Fluid Dynamics (CFD) simulations. The simulations were required by the designer either for basic checking or in case of doubts on the validity of some technical options. We present four series of simulations which lead to the detection of unsatisfactory plant behaviour, related design modification and eventually control of the variant behaviour correctness. The first three simulation series deal with the EFIT design while the forth one deals with the XT-ADS design. In the first case, the simulation put in evidence a large recirculation zone under the reactive core that had to be removed for oxygen control concern. The recirculation zone is suppressed by modifying the shape of the core support grid. In the second case, we put in evidence a recirculation zone at the entrance of the pumping system above the core. This recirculation zone can lower the pump efficiency. The entrance shape was modified to eliminate the recirculation zone. In the third case, we check the behaviour of the passive Decay Heat Removal (DHR) heat exchanger. We show that while the primary coolant flow is globally organized as expected, some flow mixing limits the efficiency of the system. The system efficiency is restored by increasing its passive pumping strength. This is performed simply extending the Heat Exchanger shroud a half-meter in the bottom direction. In the last case, we investigate the capability of an external DHR system to withstand a long complete plant shutdown. The simulation encompasses about 6 hours of physical time, enough to understand the critical trends and infers that the DHR system may be not sufficient for its purpose. This result has suggested some modification to the design (i.e. surface treatment to improve metal wall emissivity) as well as to the accident management (i.e. restart primary pumps to eliminate fluid stratification). All these design improvements have been obtained in a reasonable amount of time thanks to the continuous collaboration and exchange of information between the CFD engineer and the designer.