Showing papers in "Science and Technology of Nuclear Installations in 2012"
TL;DR: SUBCHANFLOW as discussed by the authors is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors and innovative reactors operated with gas or liquid metal as coolant.
Abstract: SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT). The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.
63 citations
TL;DR: The PANDA facility as mentioned in this paper is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests.
Abstract: The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.
41 citations
TL;DR: In this paper, a single-phase model is adopted to simulate the nanofluid convection of 1% and 4% by volume concentration, and the maximum heat transfer enhancement at the center of a subchannel formed by heated rods is ~15% for the particle volume concentration of 4% corresponding to Re = 80,000.
Abstract: Turbulent forced convection flow of Al2O3/water nanofluid in a single-bare subchannel of a typical pressurized water reactor is numerically analyzed. The single-phase model is adopted to simulate the nanofluid convection of 1% and 4% by volume concentration. The renormalization group k-e model is used to simulate turbulence in ANSYS FLUENT 12.1. Results show that the heat transfer increases with nanoparticle volume concentrations in the subchannel geometry. The highest heat transfer rates are detected, for each concentration, corresponding to the highest Reynolds number Re. The maximum heat transfer enhancement at the center of a subchannel formed by heated rods is ~15% for the particle volume concentration of 4% corresponding to Re = 80,000. The friction factor shows a reasonable agreement with the classical correlation used for such normal fluid as the Blasius formula. The result reveals that the Al2O3/water pressure drop along the subchannel increases by about 14% and 98% for volume concentrations of 1% and 4%, respectively, given Re compared to the base fluid. Coupled thermohydrodynamic and neutronic investigations are further needed to streamline the nanoparticles and to optimize their concentration.
40 citations
TL;DR: In this article, the authors present a survey of test objectives and programs carried out in the test facility at the AREVA NP in Erlangen, Germany, and describe the current state.
Abstract: Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.
36 citations
TL;DR: In this paper, an economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT), DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX), and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR).
Abstract: An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT), DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX), and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR). This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC) has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC) have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU) for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyrotechniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.
31 citations
TL;DR: The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority as mentioned in this paper, to support their emergency response team.
Abstract: Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.
30 citations
TL;DR: The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis, namely, a Radau IIA implicit Runge-Kutta method and state-of-the-art numerical solver for the first-order ordinary differential equations describing the isotope balances.
Abstract: The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, it uses a nuclear data library covering neutron- and proton-induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data, and total recoverable energies per fission. Secondly, it uses a state-of-the-art numerical solver for the first-order ordinary differential equations describing the isotope balances, namely, a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model, including such specific facilities as accelerator-driven systems. The core burn-up, activation of the structural materials, irradiation of samples, and, in addition, accumulation of spallation products in accelerator-driven systems can be calculated in a single ALEPH2 run. The code is extensively used for the neutronics design of the MYRRHA research facility which will operate in both critical and subcritical modes.
30 citations
TL;DR: The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design.
Abstract: The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.
29 citations
TL;DR: In this article, the authors developed a sensitivity and uncertainty analysis for the CASMO-4 in the context of the UAM (Uncertainty Analysis in Design, Operation, and Safety Analysis of LWRs) benchmark.
Abstract: The topic of this paper is the development of sensitivity and uncertainty analysis
capability to the reactor physics code CASMO-4 in the context of the UAM (Uncertainty Analysis
in Best-Estimate Modelling for Design, Operation, and Safety Analysis of LWRs) benchmark.
The sensitivity analysis implementation is based on generalized perturbation theory, which
enables computing the sensitivity profiles of reaction rate ratios efficiently by solving one generalized
adjoint system for each response. Both the theoretical background and the practical
guidelines for modifying a deterministic transport code to compute the generalized adjoint solutions
and sensitivity coefficients are reviewed. The implementation to CASMO-4 is described
in detail. The developed uncertainty analysis methodology is deterministic, meaning that the
uncertainties are computed based on the sensitivity profiles and covariance matrices for the
uncertain nuclear data parameters. The main conclusions related to the approach used for
creating a covariance library compatible with the cross-section libraries of CASMO-4 are presented.
Numerical results are given for a lattice physics test problem representing a BWR, and
the results are compared to the TSUNAMI-2D sequence in SCALE 6.1.
26 citations
TL;DR: In this article, the effects of upper tank geometry and water levels in the upper and lower tanks on CCFL characteristics were investigated for air-water two-phase flows at room temperature and atmospheric pressure.
Abstract: An experimental study on countercurrent flow limitation (CCFL) in vertical pipes is carried out. Effects of upper tank geometry and water levels in the upper and lower tanks on CCFL characteristics are investigated for air-water two-phase flows at room temperature and atmospheric pressure. The following conclusions are obtained: (1) CCFL characteristics for different pipe diameters are well correlated using the Kutateladze number if the tank geometry and the water levels are the same; (2) CCFL occurs at the junction between the pipe and the upper tank both for the rectangular and cylindrical tanks, and CCFL with the cylindrical tank occurs not only at the junction but also inside the pipe at high gas flow rates and small pipe diameters; (3) the flow rate of water entering into the vertical pipe at the junction to the rectangular upper tank is lower than that to the cylindrical tank because of the presence of low frequency first-mode sloshing in the rectangular tank; (4) increases in the water level in the upper tank and in the air volume in the lower tank increase water penetration into the pipe, and therefore, they mitigate the flow limitation.
25 citations
TL;DR: In this article, the interfacial behavior during countercurrent two-phase flow of air-water and steam-water in a model of a PWR hot leg was studied quantitatively using digital image processing of a subsequent recorded video images of the experimental series obtained from the TOPFLOW facility, Helmholtz-Zentrum Dresden Rossendorf e.V. (HZDR), Dresden, Germany.
Abstract: The interfacial behavior during countercurrent two-phase flow of air-water and steam-water in a model of a PWR hot leg was studied quantitatively using digital image processing of a subsequent recorded video images of the experimental series obtained from the TOPFLOW facility, Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR), Dresden, Germany. The developed image processing technique provides the transient data of water level inside the hot leg channel up to flooding condition. In this technique, the filters such as median and Gaussian were used to eliminate the drops and the bubbles from the interface and the wall of the test section. A Statistical treatment (average, standard deviation, and probability distribution function (PDF)) of the obtained water level data was carried out also to identify the flow behaviors. The obtained data are characterized by a high resolution in space and time, which makes them suitable for the development and validation of CFD-grade closure models, for example, for two-fluid model. This information is essential also for the development of mechanistic modeling on the relating phenomenon. It was clarified that the local water level at the crest of the hydraulic jump is strongly affected by the liquid properties.
TL;DR: In this article, the optimal position for placing a detector near a very intense source is defined as the one with the minimal relative uncertainty in the counting rate, which can be applied with an accelerator-driven subcritical system or a spallation source.
Abstract: The operation of accelerator-driven systems or spallation sources requires the monitoring of intense neutron fluxes, which may be billions-fold more intense than the fluxes obtained with usual radioactive sources. If a neutron detector is placed near a very intense source, it can become saturated because of detector dead time. On the contrary, if it is placed far away from the source, it will lose counting statistics. For this reason, there must exist an optimal position for placing the detector. The optimal position is defined as the one with the minimal relative uncertainty in the counting rate. In this work, we review the techniques to determine the detector dead time that can be applied with an accelerator-driven subcritical system or a spallation source. For the case of a spallation source, counting rates do not follow Poisson's statistics because of the multiplicity of the number of neutrons emitted by incident proton. It has been found a simple expression that relates the optimal counting rate with the source multiplicity and the uncertainty in the determination of the dead time.
TL;DR: In this article, the authors present a review of the main characteristics of the experimental facility, analyze the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility.
Abstract: Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.
TL;DR: In this paper, the authors investigated the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding, and the results indicated that the combined use of and reprocessed fuel allowed production without the initial requirement of enrichment.
Abstract: Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was for production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of without the initial requirement of enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor () evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of and reprocessed fuel allowed production without the initial requirement of enrichment.
TL;DR: In this article, the authors proposed a generalized heat transfer coefficient (HTC) based on curve fitting of most of the reported semi-empirical condensation models, which are valid for specific wall conditions.
Abstract: In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.
TL;DR: Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.
Abstract: For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.
TL;DR: Fuzzy-logic-based diagnostic monitoring is applied to data acquired from instrumentation within operating facilities and the decision making regarding the system status and the possible need for component maintenance is performed.
Abstract: Economic constraints are driving the electric power industry to seek improved methods for monitoring, control, and diagnostics. To increase plant availability, various techniques have been implemented in industry to assess equipment condition to prevent system inoperability. The availability of a large number of measured signals and additional component information and the increasing number of signal processing options to analyze sampled data motivate the assimilation of such diverse information into a plantwide condition monitor. The use of fuzzy logic is described herein for the purpose of performing the decision making regarding the system status and the possible need for component maintenance. Fuzzy-logic-based diagnostic monitoring is applied to data acquired from instrumentation within operating facilities.
TL;DR: In this article, a three-dimensional computational fluid dynamics (CFD) method was used to model the boiling two-phase flow in one of the PSBT 5-by-5 rod bundle tests.
Abstract: Three-dimensional computational fluid dynamics (CFD) method was used to model the boiling two-phase flow in one of the PSBT 5-by-5 rod bundle tests. The rod bundle with all the spacers was modeled explicitly using unstructured computational grids. The six-equation, two-fluid model with the wall boiling model was used to model the boiling two-phase flows in the bundle. The computed void fractions compare well with the measured data at the measuring plane. In addition to the averaged void data, the CFD results give a very detailed picture of the flow and void distributions in the bundle and how they are affected by solid structures in the flow paths such as the spacer grids and mixing vanes.
TL;DR: In this article, the reaction rates of activation foils set in the core region and at the location of the target are measured by the foil activation method to obtain the subcritical multiplication parameters.
Abstract: Basic experiments on the accelerator-driven system (ADS) at the Kyoto University Critical Assembly are carried out by combining a solid-moderated and -reflected core with the fixed-field alternating gradient accelerator. The reaction rates are measured by the foil activation method to obtain the subcritical multiplication parameters. The numerical calculations are conducted with the use of MCNPX and JENDL/HE-2007 to evaluate the reaction rates of activation foils set in the core region and at the location of the target. Here, a comparison between the measured and calculated eigenvalues reveals a relative difference of around 10% in C/E values. A special mention is made of the fact that the reaction rate analyses in the subcritical systems demonstrate apparently the actual effect of moving the tungsten target into the core on neutron multiplication. A series of further ADS experiments with 100 MeV protons needs to be carried out to evaluate the accuracy of subcritical multiplication parameters.
TL;DR: In this article, an evaluation of the Fukushima-Daiichi nuclear power plant (NPP) accident in Units 1 to 4 is presented, where the authors make an attempt to discuss the scenario within a technological framework, considering precursory documented regulations and predictable system performance.
Abstract: The paper deals with the evaluation of the Fukushima-Daiichi Nuclear Power Plant (NPP) accident in Units 1 to 4: an attempt is made to discuss the scenario within a technological framework, considering precursory documented regulations and predictable system performance. An outline is given at first of the NPP layout and of the sequence of major events. Then, plausible time evolutions of relevant quantities in the different Units, is inferred based on results from the application of numerical codes. Scenarios happening in the primary circuit and containment (three Units involved) are distinguished from scenarios in spent fuel pool (four Units involved). Radiological releases to the environment and doses are approximately estimated. The event is originated by a natural catastrophe with almost simultaneous occurrence of earthquake and tsunami. These caused heavy destruction in a region in Japan much wider than the land around the NPP which was affected by the nuclear contamination. Key outcome from the work is the demonstration of strength for nuclear technology; looking at the past, misleading Probabilistic Safety Assessment (PSA) data and inadequacy in licensing processes have been found. Looking into the future keywords are Emergency Rescue Team (ERT), Enhanced Human Performance (EHP), and Robotics in Nuclear Safety and Security (RNSS).
TL;DR: In this article, the authors measured the characteristics of countercurrent flow limitation (CCFL) in a 1/10-scale model of the surge line using air and water at atmospheric pressure and room temperature.
Abstract: Steam generated in a reactor core and water condensed in a pressurizer form a countercurrent flow in a surge line between a hot leg and the pressurizer during reflux cooling. Characteristics of countercurrent flow limitation (CCFL) in a 1/10-scale model of the surge line were measured using air and water at atmospheric pressure and room temperature. The experimental results show that CCFL takes place at three different locations, that is, at the upper junction, in the surge line, and at the lower junction, and its characteristics are governed by the most dominating flow limitation among the three. Effects of inclination angle and elbows of the surge line on CCFL characteristics were also investigated experimentally. The effects of inclination angle on CCFL depend on the flow direction, that is, the effect is large for the nearly horizontal flow and small for the vertical flow at the upper junction. The presence of elbows increases the flow limitation in the surge line, whereas the flow limitations at the upper and lower junctions do not depend on the presence of elbows.
TL;DR: A review of the methods used is provided in this article, and several examples illustrate the success of the methodology in reactor physics and a new application as the improvement of nuclear basic parameters using integral experiments is also described.
Abstract: The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.
TL;DR: A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation), has been operated by KAERI as mentioned in this paper.
Abstract: A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.
TL;DR: The SARNET2 project as mentioned in this paper was the first project in the 6th Framework Programme (FP6) of the European Commission, coordinated by IRSN, for 4 years in the FP7 frame.
Abstract: Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.
TL;DR: In this paper, the safety requirements for a large-scale Japan sodium-cooled fast-reactor (JSFR) conformed to the defense-in-depth principle in IAEA.
Abstract: As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.
TL;DR: In this article, the authors performed a station black-out analysis for Atucha 2 nuclear power plant in Argentina, where the first safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total.
Abstract: A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.
TL;DR: In this article, a loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed.
Abstract: The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.
TL;DR: The PWR PACTEL as discussed by the authors test facility was designed to model the thermal-hydraulic behavior of VVER-440-type nuclear power plant (EPR) type ReF.
Abstract: This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.
TL;DR: In this paper, the improvement of in-service inspection and repair (ISI&R) capabilities was identified as a major issue within the French-associated multiannual SFR research program, the periodic examination and repair are looked at through the following main R&D axes: (i) improvement of the primary system conceptual design in order to ease periodic examination, (ii) development of inspection techniques, (iii) accessibility and associated robotics, and (iv) development and validation of repair processes.
Abstract: In the frame of the CEA, EDF, and AREVA coordinated research program launched in 2007 for the development of Generation IV sodium-cooled fast reactors (SFRs), the improvement of in-service inspection and repair (ISI&R) capabilities was identified as a major issue. Within the French-associated multiannual SFR research program, the periodic examination and repair are looked at through the following main R&D axes: (i) improvement of the primary system conceptual design in order to ease periodic examination and repair, (ii) development of inspection techniques (periodic inspection tools and associated simulation), (iii) accessibility and associated robotics, and (iv) development and validation of repair processes. Associated needs are being defined through an iterative method between designers and inspection specialists: adaptation of the SFR design to ISI&R requirements, validation of the ultrasonic transducers, of associated ultrasonic nondestructive examination techniques, of laser repair processes, of associated robotic equipment. International collaboration is also running for some specific items such as ultrasonic visualization under liquid sodium.