N. Muellner

NUTEMA: a tool for supervising nuclear technology and for the transfer of knowledge

In this paper is presented NUTEMA (Nuclear Power Plant Technology Management System) an IT innovative system dealing with the nuclear technology. The overall goal is to offer a framework capable to present in one room whatever is concerned with a Nuclear Power Plant (NPP) unit implementing together a multilevel and multi-view user access according to enterprise architecture models. The NUTEMA system assembles all together the competences needed for the construction, the design, the management of an NPP, and allows the training of nuclear industry staff and stakeholders. The concept of IAEA Design Authority has been the starting idea at the base of the NUTEMA design. It is primarily devoted to utilities, but any industry like regulatory authorities and research institutes concerned with nuclear technology can take benefits from its use.

The Best-Estimate Plus Uncertainty (BEPU) Challenge in the Licensing of Current Generation of Reactors

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.NPP

BEPU Approach in Licensing Framework, Including 3D NK Applications

During the recent years, a world-wide renewed interest in the exploitation of nuclear energy for electricity production is seen among both the Western and the new industrializing Countries (e.g., China and India). As a result, 61 reactors are now under construction and more than 100 units are planned for the incoming decade. Such impressive development is totally based on Light and Heavy Water Reactor (LWR & HWR) technologies [1], on designs that are an evolution of the robust and reliable Nuclear Power Plants (NPP) designed and built during the seventies-eighties of the last century. At that time, the need to guarantee an high safety level on one side and on the other the limited computational capabilities and the scarce knowledge of some phenomena, drove the main nuclear safety authorities to establish extremely conservative rules. Nowadays, after that tremendous progress has been made into the computational power availability, models accuracy and the knowledge of relevant phenomena, there is the need to go toward more realistic safety analyses and to relax some levels of conservativeness without compromising the always elevated safety level of the nuclear industry. The aim of this Chapter is to give an overview of the current trends in the licensing frameworks for a NPP. International best-practices are presented and discussed and sample applications derived from works of the San Piero a Grado Nuclear Research Group of the University of Pisa (GRNSPG/UNIPI) on existing industrial facilities are also reported.

SCALING ANALYSIS IN BEPU LICENSING OF LWR

“Scaling” plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a “roadmap” to scaling. Key elements are the “scaling-pyramid”, related “scaling bridges” and a logical path across scaling achievements (which constitute the “scaling puzzle”). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the “key-to-scaling”, in the licensing process of a nuclear power plant. The proposed “road map to scaling” aims at solving the “scaling puzzle”, by introducing a unified approach to the problem.

Addressing Boron Dilution Scenario Through RELAP5/3.3 Analysis of PWR SB LOCA

Reactivity accident scenarios can occur originated by internal boron dilution in the primary system of a nuclear pressurized water reactor type (PWR or VVER). In essence the problem is caused by boron dilution following vaporization and condensation of the primary system coolant in case of decrease of primary system mass inventory, for example during a small-break loss of coolant accident (SB-LOCA) that may include boiling in the core with condensation of steam in the steam generators. When the liquid level in the reactor vessel decreases below the hot leg elevation, steam begins to flow to the steam generators and condenses there. This steam carries no boron and thus boron concentration in the cold leg loop seals begins to decrease. If for some reason this water plug with low boron concentration begins to flow towards the core and enters it without any major mixing with the borated coolant, the result is a positive reactivity insertion. The paper presents an analysis by RELAP5 Mod 3.3 code [1] of a small break LOCA of 20 cm2 area in the lower plenum of a four-loop PWR nuclear reactor. The boundary conditions of the calculations consider the eight accumulator tanks available, two/four low pressure injection systems (LPIS) available, and two of the four high pressure injection systems (HPIS) available. Sensitivity calculations were performed, regarding among other things, the boron concentration in the Emergency Core Cooling Systems (ECCS) and reactor cooling system (RCS) from Design Basis Accident (DBA) to beyond DBA conditions. From the results obtained, in some calculations boron dilution is observed in more than one loop seal. The situation in which the plugs in the loop seals are transported to the core without mixing with other borated water led to a potentially hazardous situation for four calculations in which initial conditions were far from DBA. It is important to emphasize that the present study has not the objective of a safety analysis of the NPP involved, but it should be considered inside research activities regarding the boron dilution issue.Copyright

Optimizing the Initial Pressure of Accumulators for the Atucha2 NPP Using an Optimization Method

Passive safety systems like hydro accumulators offer high reliability and are therefore, when a choice is possible, often preferred over active safety systems. However, their effectiveness in case of an incident or accident depends on a large number of parameters (like break size in case of a loss of coolant accident, availability of other safety systems, initial pressure in the accumulators) and is in general difficult to predict. This paper presents a study to optimize the initial pressure and the pressure drops in the accumulator line for a intermediate break loss of coolant accident for Atucha 2, a Siemens-KWU, heavy water moderated, channel type pressurized water reactor under construction. An optimization method was applied. The thermal hydraulic system code RELAP5 mod 3.3 was used for the analysis. Three cases have been analyzed. First, the initial pressure and pressure drop in the accumulator line was optimized in case of an intermediate break in cold leg two, assuming safety injection of two of the four trains of safety systems into hot and cold leg. Second, like before, but assuming safety injection into cold leg only. Third, like case two, but grouping the four accumulators in two groups, with different initial pressure and pressure drops in the accumulator lines. The results show that a slight increase of the initial accumulator pressure compared to the design value could be beneficial for the investigated initial event. Further, case three shows that different initial pressure in the accumulators could increase the effectiveness of the intervention for the investigated accident.Copyright

Accident Management in VVER-1000

The present paper deals with the investigation study on accident management in VVER-1000 reactor type conducted in the framework of a European Commission funded project. The mentioned study involved both experimental and computational fields. The purpose of this paper is to summarize the main findings from the execution of a wide-range analysis focused on AM in VVER-1000 with main regard to the qualification of computational tools and the proposal for an optimal AM strategy for this kind of NPP.

Investigation of a Possible Emergency Procedure for the VVER 1000 NPP in Case of a Total Loss of Feedwater and a Main Steam Line Break

Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper “A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario” that it is possible to safely control a total loss of feedwater scenario in the Angra2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, i.e. long term core cooling is achievable.© 2004 ASME