Conjugating ALARA, BEPU, Safety Margins and Independent Assessment in Nuclear Reactor Safety

ALARA (As-Low-As-Reasonably-Achievable) is an early principle in Nuclear Reactor Safety, NRS (Nuclear Reactor Safety): Designers and Operators must do their best to minimize doses to the humans. BEPU (Best Estimate Plus Uncertainty) is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU, or ALARA is at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margin (SM) is an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The SM concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit. Here the environment includes the personnel in charge of activities connected with the NPP. The Extended SM concept, E-SM, implies the formulation of some ten-thousands SM values, which shall correspond to a similar number of monitored variables. Reasons for E-SM are the examples in section 4.1. Independent Assessment (IA) is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. In the paper we discuss that: a) ALARA is at the origin of BEPU; b) BEPU-FSAR analyses are the natural origin of E-SM values; c) The implementation of E-SM equals to introducing an additional physical barrier against the release of fission products

A view on identification of thermal-hydraulic phenomena for validation of best-estimate computer codes

Validation of best-estimate codes is a necessary step to prove their applicability to calculate accident scenarios, including the course of events. It shall demonstrate that those physical phenomena, which are important for a scenario, are calculated appropriately. A common list of 113 thermal-hydraulic phenomena is provided, based on previous reports of OECD/NEA-CSNI and IAEA, including Separate Effects Test (SET) facilities and Integral Test facilities of PWRs and BWRs, VVERs, Advanced Reactors, as well as Containment. Added objective of the activity is to show that the list of phenomena is applicable to the major number of water-cooled reactors. Twelve reactor types are considered for the characterization of 47 accident scenarios cross-linked with the identified phenomena. Focus of this paper is on the updated identification of the list of thermal-hydraulic phenomena

Benchmark of Atucha-2 PHWR RELAP5-3D Control Rod Model by Monte Carlo MCNP5 Core Calculation

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3DC/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2

An Integrated Software Platform for Best Estimate Safety Analyses of Nuclear Power Plants

Nuclear power plant safety is granted through the demonstration that regulatory acceptance criteria are fulfilled by the provided (calculated) analyses of the NPP performances and sufficient safety margins are respected during normal operation, anticipated transients and postulated accident conditions. Safety margins are very hard to determine in absolute terms, numerical calculations are used to assess their values. Over the last 30 years an extensive effort has been carried out aiming to improve the knowledge of the nuclear power plant behaviour under transient scenarios. The development of Best Estimate (BE) computer codes are the direct consequence of these noteworthy efforts. The availability of more sophisticated and specialized computer codes gives the analyst the possibility to perform very detailed analysis in all the fields involved in the safety of a NPP: thermal-hydraulics, CFD, 3D neutron kinetics etc. The possibility to create a software environment where a multidisciplinary problem can be solved adopting different specialized codes able to exchange data among them is a fruitful approach to the problem aiming to improve the results. The computational tools, adopted in best-estimate approach in licensing, include a) the best estimate computer codes; b) the nodalizations together with the procedures for the development and the qualification; c) the uncertainty methodology. The Nuclear Research Group of San Piero a Grado of the University of Pisa has developed a software platform with 15 interacting computer codes. Such platform covers the reactor simulation multidisciplinary problem from generation of neutron cross-sections, through system thermal-hydraulic analyses, up to detailed structural and fuel mechanics studies and it embeds software procedures for automatized data transfer between codes. Together with methodological procedures for nodalizations development and qualification the platform leads to a great decrease of the human induced error in the results. The developed platform has been tested and successfully applied to perform the safety analyses required by the Chapter 15 of the Final Safety Analysis Report of the CNA-2 nuclear power plant in Argentina

An overview of Thorium Utilization in Nuclear Reactors and Fuel Cycle

The Nuclear Power Plants (NPP) constructed in the XX century, also called generation II reactors, are still in operation, most of them Light Water Reactors, but are being decommissioned. These reactors have a low burn up (~30 MWD/kg U) and utilize UO2 as nuclear fuel and are operating in a Once Through Cycle (OTC); they use a very low energy content of the natural resources (~0,5%). To overcome economic and political and partly safety issues, since the end of last century, and beginning of this century, the nuclear industry launched a new generation of evolutionary reactors, called Generation III, such as the Westinghouse AP 1000, and AREVA EPR. These reactors still use uranium as primary source but have an increased burn up (~60 MWD/Kg U), which although increasing the utilization of the natural resources (up to 1%), still are not significant to be considered sustainable: if only uranium is used in an OTC, uranium will be exhausted in this century. To increase the utilization of natural resources, recycling of uranium and plutonium is already in use in many countries and used as Mixed Oxide of U-Pu fuel (MOX) in the same thermal reactors. To turn nuclear energy sustainable, a long-term deployment of innovative reactors is underway. These reactors and their associated fuel cycle are old concepts with technological improvements and generically denominated as Generation IV, are in development and, in some cases, they are breeders, HLW burners, and efficient concepts. Another concept that although not new is constitute by the Small Modular Reactors (SMR), with power less than 300 MWe, which nowadays are deserving a lot of attention by the nuclear industry. Another option is to utilize thorium as a primary source of energy. Although not fissile at thermal energy, it produces 233U, which is one of best fissile nuclide (number of neutrons produced per neutron absorbed). Also, it is three times more abundant than uranium in the earth crust and has thermal physics properties when used as (U-Th) O2 better than UO2. Several Th/U fuel cycles, using thermal and fast reactors were proposed and are still under investigation. Although, the first reactors to utilize thorium were PWR, using (U-Th)O2, such as the Indian Point, and Shipping Port, thorium has been proposed as fuel for the molten salt reactor, the advanced heavy water reactor, High Temperature Reactors, Pebble Bed reactor, fast breeder reactors, and more recently, for the innovative accelerator driven system in a double strata fuel cycle and for the Generation IV, such as the LFTR - Liquid Fluoride Thorium Reactor, which is a self-sustainable Molten Salt Reactor, promising to turn nuclear energy by fission in a sustainable source, with a utilization of the natural resources of 100%. This paper, besides an introduction of the present time uranium fuel cycles, will give an over view of the thorium utilization in nuclear reactors and fuel cycles, with an emphasis in Advanced PWR

Identification of Limiting Case Between DBA and SBDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D© system code. Within the framework of the third Agreement “NA-SA – University of Pisa” a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions

Comparative study between cold leg and hot leg safety injection during SBLOCA in a 4-loop PWR NPP

This article presents a comparison between two operation modes for the emergency core cooling system during a Small Break Loss of Coolant Accident (SBLOCA) in the cold leg of 4-loop PWR Westinghouse design nuclear power plant. In the first mode, the cold leg safety injection is used to mitigate the consequences of the accident and in the second mode the hot leg safety injection is used. The best estimate light water reactor transient analysis system code RELAP5 Mod3.3 was used in calculations. The plant nodalization consists of two loops; the first one represents the broken loop and the second one represents the other three intact loops. The results show that, in the cold leg safety injection the primary pressure decreases with time and remains higher than the secondary pressure for a period of time (~ 500 sec) during whichthe steam generators remains as a heat sink for the primary side, the accumulators start late and functioning on remaining transient time, and a repeatable loop seal clearing and refill occurs. During the hot leg safety injection the primary pressure decreases rapidly but remains higher than the secondary pressure for a longer period of time (~ 600 sec), the accumulators start early and functioning on a part of the transient time before they are totally discharged, and there is no repeatable loop seal clearing and refill. In the two modes the maximum clad surface temperature does not violate the safety limit

The April-May 2006 volcano-tectonic events at Stromboli volcano (Southern Italy) and their relation with the magmatic system.

Between April 10th and May 22th 2006, a small seismic swarm of 5 volcano-tectonic events occurred on the volcanic island of Stromboli (Southern Italy). Two of these, having M > 3 and an intensity of about V-VI MCS, were clearly felt causing concern in the population. They were recorded during a period of increased explosive activity and were followed by two major explosions at the summit craters on May 22th, few hours after the last earthquake and on 16th June. The location of such events has been performed using a probabilistic approach based on the Equal Differential Time tecnique. Using this tecnique, we were able to locate all the events, showing how they cluster below the volcanic edifice at a depth of about 5÷6 km. From observed P wave polarities we determined the focal mechanisms of the 4 major events. Using earthquake scaling nlaws, we calculated the fault area and the average slip for the two major events. Finally, assuming an homogeneous half-space model we computed the isotropic stress changes below the volcano edifice. The negative stress variation over the central axis of the volcano suggests that the earthquakes were triggered by a pressurization of the magmatic system

Enhanced Nuclear Engineering Simulators

Engineering simulation is a sophisticated multi-purpose technology allowing the users of simulators to run a variety of engineering activities due to the possibility of modifying the simulated plant architecture and components, to adjust parameters, to test alternative solutions. Engineering Simulators (ES) have been built and used worldwide for a variety of purposes: - Development and refinement of the plant design or plant modifications - Safety analyses focused on the overall system behaviour - Verification and Validation (V&V) of systems and components - Development of Operational and Emergency Procedures - Pre-Training of operators and supervisors - High level education and Communication activities - Human Factor Engineering Analysis - Adaptive Control System training Engineering Simulators also play a role in developing and maintaining key nuclear skills, as knowledge repositories and tools for training at various levels of expertise

AUTOMATIC ANALYSIS OF SEISMIC DATA BY USING NEURAL NETWORKS: APPLICATIONS TO ITALIAN VOLCANOES.

The availability of the new computing techniques allows to perform advanced analysis in near real time, improving the seismological monitoring systems, which can extract more significant information from the raw data in a really short time. However, the correct identification of the events remains a critical aspect for the reliability of near real time automatic analysis. We approach this problem by using Neural Networks (NN) for discriminating among the seismic signals recorded in the Neapolitan volcanic area (Vesuvius, Phlegraean Fields). The proposed neural techniques have been also applied to other sets of seismic data recorded in Stromboli volcano. The obtained results are very encouraging, giving 100% of correct classification for some transient signals recorded at Vesuvius and allowing the clustering of the large dataset of VLP events recorded at Stromboli volcano