the best estimate plus uncertainty challenge in the current licensing process of present reactors

Within the licensing process of the KWU Atucha II PHWR (Pressurized Heavy Water Reactor), the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: (a) the selection of PIE (Postulated Initiating Events) and (b) the analysis by best estimate models supported by uncertainty evaluation. Otherwise, key elements of the approach are (1) availability of qualified computational tools including suitable uncertainty method, (2) demonstration of quality, and (3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper, and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified

The Best-Estimate Plus Uncertainty (BEPU) Challenge in the Licensing of Nuclear Power Plants (NPP)

Within the licensing process of the Atucha II PHWR (Pressurized Heavy Water Reactor) the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: a) the selection of PIE (Postulated Initiating Events) and, b) the analysis by best estimate models supported by uncertainty evaluation. Otherwise, key elements of the approach are: 1) availability of qualified computational tools including suitable uncertainty method; 2) demonstration of quality; 3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified

The BEPU Challenge in Current Licensing of Nuclear Power Reactors

Several ways can be adopted to define BEPU, one of this being “BEPU connection between SYS TH code development and V&V, Scaling, Uncertainty on the one side and licensing process on the other side.” Then, “BEPU is the application of SYS TH codes.” The BEPU process shall be associated with the licensing process. The licensing process noticeably includes the PSA together with a number of methods not discussed in the present book; those methods including PSA need a cross-link with BEPU. AA is part of the licensing process. Among the general attributes of AA, the first one shall be the compliance with the established regulatory requirements. The second attribute deals with the adequacy and the completeness of the selected spectrum of events. The achievement of the spectrum of events (or envelope) shall be the result of the combined applications of deterministic and probabilistic methods. The third attribute is connected with the knowledge base including that captured by the qualified computational tools and analytical procedures suitable for the analysis of transient conditions envisaged for individual (concerned) NPP. The complexity of a NPP and/or the accident scenarios may prove challenging for a conservative analysis, thus justifying the choice for a BEPU approach. This implies two main needs for nuclear thermal-hydraulics: (a) to adopt the current computational tools, proving (to the regulatory authority) an adequate quality via suitable V&V and (b) to adopt a qualified uncertainty method

Atucha-I NPP Containment Sensitivity Calculations to Support Uncertainty Evaluations

A key part of the safety framework of a Nuclear Power Plant (NPP) is represented by the containment behaviour. The GOTHIC thermal hydraulic code has been employed for evaluating the Atucha-I NPP containment responses during two postulated severe accident scenarios, Station Black Out and Large Break Loss of Coolant Accident without Safety Injection Pumps, while assuming that the external cooling of the Reactor Pressure Vessel is carried out during the transients. The target of the analyses is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the external vessel cooling. This could lead to pressure values above the safety limit. The containment pressure, temperature, the distribution of hydrogen and the liquid level have been analysed as target variables. Two different nodalizations were used, a “detailed” nodalization, meant to have the most refined according to the available computational resources, and a “coarse” nodalization in order to lower the computational requirements, without significantly compromising the global response. Several sensitivities were simulated for 100,000 seconds and were meant to understand and characterize the impact of the different nodalization parameters (geometrical aspects, material properties, Boundary Conditions) and as such also provide some indications of some important uncertainty parameters contribution

Identification of Limiting Case Between DBA and Selected BDBA (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 RELAP5-3DC system code. Within the framework of the third Agreement 'NA-SA - Univ. 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

A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper

Atucha-1 NPP containment venting analysis following SBO and LBLOCA events by GOTHIC code

Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant (NPP). The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated severe accident scenarios, Station Black Out and Large Break Loss of Coolant Accident without Safety Injection Pumps (SIPs), while assuming that the external cooling of the Reactor Pressure Vessel is carried out during the transients. The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between. The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the external vessel cooling, which could cause an increase in containment pressure, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been analysed as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The “detailed” nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number of cells. Taking into consideration that several sensitivities were performed, the “coarse” nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections. Both accident transients, for each type of nodalization, were simulated for 200,000 s. At the end of the simulated transient, results showed that for the Large Break Loss of Coolant Accident pressure is predicted to surpass 5 bar, while the Station Black Out scenario is calculated to reach 4.4 bar. The performed sensitivities were simulated for 100,000 s and were meant to understand and characterize the impact of the different nodalization parameters (geometrical aspects, material properties, BCs). In addition, due to several code anomalies identified, several other sensitivity calculations were performed in order to find a way to analyse and mitigate the issues

Optimizing the Initial Pressure of Accumulators for the Atucha-2 Nuclear Power Plant using an Optimization Method

Accumulators (ACC) constitute passive systems essential part of the Emergency Core Cooling Systems (ECCS) of any water cooled and moderated reactor. Key design parameters for the ACC are the pressure and the volume; in this last case liquid and gas volumes are distinguished. In relation to pressure: a high initial value brings to early ACC intervention in case of Loss of Coolant Accident (LOCA), increasing the ACC fluid mass lost to the break; a low initial value brings to late actuation and danger of high rod surface temperature at the time of actuation. The optimized design value of initial pressure is more difficult to be attained in case of Small Break LOCA because of the several possible accident scenarios. A procedure based on ‘analytical optimization methods’ has been set up and tested to identify the best initial pressure for the Atucha-II Nuclear Power Plant under final stage of construction in Argentina

ATUCHA-1 NPP CONTAINMENT VENTING ANALYSIS FOLLOWING SBO AND LBLOCA EVENTS BY GOTHIC CODE

Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant. The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated accident scenarios, Station Black Out and Large Break Loss of Coolant Accident, while assuming the external cooling of the Reactor Pressure Vessel is carried out during the transients. The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between. The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the ex-vessel cooling, which could cause an excessive pressurization of the containment, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been monitored as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The "detailed" nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number cells. While the "coarse" nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections, e.g. all doors and blow off panels have been simulated to open with the designed differential pressure logic. Both accident transients, for each type of nodalization, were simulated for 200,000 seconds. Results showed that for the Large Break Loss of Coolant Accident pressure is predicted to reach around 5.25 bar, while the Station Black Out Scenario reaches 4.4 bar