1,721,191 research outputs found
Results of RELAP4/MOD6 code applications", Lectures L8, L9, L10, L13, L16 at Course on Thermal-hydraulic Phenomena in Nuclear Reactor Technology - Sofia (BG),
No full textA Course in Nuclear thermal-hydraulics was organized by UNIPI in Sofia at the time of the cold war (Soviet Union collapsed in 1991). Contacts crossing the iron curtain were extremely complex. The entire Course consists of several hundred slides (all preserved in paper format by the corresponding author) and a couple dozen lectures (see copies below). Two pages from the Course are reported below. The current lecture deals with the selected topics relevant in nuclear thermal-hydraulics –more details can be found in the copied program below
"Analysis of natural circulation test A2-77 performed in LOBI facility, by RELAP5/MOD2 code"
No full textThe document deals with the analysis of an experiment to be performed in LOBI ITF in the configuration Mode2. LOBI was a Pressurized Water Reactor (PWR) simulator installed in the EC center of Ispra Italy financially supported by EURATOM.
The configuration Mod2 of LOBI was designed as a consequence of the Three Mile accident in US and is the follow-up of the configuration Mod1. Mod2 was designed to simulate stratification phenomena and natural circulation which are important during Small Break Loss of Coolant Accidents (SBLOCA); as a difference Mod1 was designed to simulate phenomena expected during Large Break Loss of Coolant (LBLOCA) scenarios. Namely downcomer size in Mod2 was much smaller than in the case of Mod1 and a large number of Emergency Core Cooling Systems (ECCS) were installed in Mod2. The concerned A2-77 Natural Circulation (NC) experiment was performed to model as far as possible the PWR performance. Various phase of NC were identified and brought to important findings in Nuclear Reactor Safety, including the characterization of instabilities and the reflux condensing mode.
The work was discussed in a specific meeting in Ispra of the so-called LPTF research group where University of Pisa represented Italy
Accuracy quantification by the FFT method in FARO L-14 (ISP 39) open calculations, Ispra (I), April 22-23, 1997
No full textThe Fast Fourier Transform Based Method (FFTBM) was developed at University of Pisa to achieve a quantitative evaluation of the accuracy of thermal-hydraulic system code calculations. International cooperation was established to transfer the method with various Institutions all over the world. In the case documented in this report one scientists from University of Pisa (Mario Leonardi) has been invited at the European Commission (EC) Joint Research Center (JRC) of Ispra (Varese, Italy) to implement and to apply the method. The present document has been issued by JSI and describes the application of the method to the severe accident (corium interaction with water) FARO L-14 carried out in the framework of EC EURATOM researches. The L-14 experiment was selected at the basis of International Standard Problem 39 (ISP-39) by the Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Installations (OECD/NEA/CSNI)
Best Estimate Analysis and Uncertainty evaluation of the Angra-2 LBLOCA DBA
No full textThe licensing is a legal process to transfer the nuclear safety technology (developed and supported by researches) into a legal-fixed environment. Competences in the legal nuclear safety framework (typically part of the Atomic Act in each Country) and of nuclear reactor safety are needed in such a context.
University of Pisa had the opportunity and the challenge to participate, on the behalf of the local Regulatory body (a Governmental Institution) in Brazil, to the review of the safety report submitted by the German Industry (namely, KWU-Siemens) for the licensing of the Angra-2, 1400 Mwe (the largest Nuclear Power plant in the world in terms of reactor power) o Nuclear Power Plant.
The present document deals with the independent analysis of the Large Break Loss of Coolant Accident (LBLOCA) performed at University of Pisa and to be compared with the results for the same accident obtained by the industry
"Post-test analysis of LOBI experiment BL-21 by RELAP5/MOD2 code", 9th Meet. of LOBI LPTF, Ispra (I), Sept. 26-28, 1988
No full textThe document deals with the post-test of an experiment to be performed in LOBI ITF in the configuration Mode2. LOBI was a Pressurized Water Reactor (PWR) simulator installed in the EC center of Ispra Italy financially supported by EURATOM.
The configuration Mod2 of LOBI was designed as a consequence of the Three Mile accident in US and is the follow-up of the configuration Mod1. Mod2 was designed to simulate stratification phenomena and natural circulation which are important during Small Break Loss of Coolant Accidents (SBLOCA); as a difference Mod1 was designed to simulate phenomena expected during Large Break Loss of Coolant (LBLOCA) scenarios. Namely downcomer size in Mod2 was much smaller than in the case of Mod1 and a large number of Emergency Core Cooling Systems (ECCS) were installed in Mod2. The concerned BL-21 Steam Generator Tube Rupture (SGTR) experiment was performed to model as far as possible the PWR performance. The code application allowed to confirm the scaling criteria adopted for the design of the test. The SGTR is a safety system challenging situation where Emergency Operating Procedures must be carefully designed to avoid or minimize radioactivity release to the environment, avoid the risk of boron dilution and Pressurized Thermal Shock and maintain cooled the core.
The present post-test analysis, follow-up of works to plan the experiment and to perform pre-test analysis was discussed in a specific meeting in Ispra of the so-called LPTF research group where University of Pisa represented Italy
Outline of User effect on Codes Predictions, 7th Meet. of CSNI THSB Task Group, Paris (F), June 26-28, 1990
No full textThe execution of a large number of pre-test and post-test analyses concerning (not only) International Standard Problems (ISP) each one with several participants with many of them using the same code version, showed an important impact on the code prediction which is not (or better, not entirely) associated with the nature of the code. This was called user effect. The user effect became a dominating topic in discussions and plant to improve the technology. The present report constitutes the origin of the ‘user-effect’ wording
Experience in the assessment of RELAP5/MOD2 code at Pisa University ICAP Meet., Grenoble (F), March 1-4, 1988
No full textA variety of experiments performed in Integral Test Facilities (ITF), like LOBI, LOFT, SEMISCALE, etc., were at the basis of the demonstration of capabilities and deficiencies of the Relp5 code. Both code deficiencies and capabilities were discussed at the International Code Assessment Program (ICAP) managed by United States Nuclear Regulatory Commission (US NRC). The opinion by UNIPI was presented in a Workshop held in Grenoble
Relevant results obtained in the analysis of LOBI/MOD2 natural circulation experiment A2-77A
No full textThe analysis of LOBI experiment A2-77A Natural Circulation test is documented. Full document is attache
Best Estimate Plus Uncertainty Approach in nuclear reactor safety and licensing: brief history and elements
Full text linkThe best es ti mate plus un cer tainty is, at the same time, an ap proach, a pro ce dure and a frame - work in nuclear thermal-hydrau lics and nuclear reactor safety and licensing. The motiva tion at the ba sis of the best es ti mate plus un cer tainty is the lack of knowl edge in the ar eas of sin gle and, mainly, two-phase tran sient ther mal-hy drau lics. In other terms and in tro duc ing some simplifications, the insuf ficient knowledge of tur bulence imposes the design of roadmaps for the application of imperfect (thermal-hydraulic) models to the evaluation of design features and of safety for com plex tech no log i cal in stal la tions or sys tems like the Nu clear Power Plants and, more specif ically, the water cooled nuclear reac tors. Furthermore, the legal counter part of nu clear re ac tor safety, or the li cens ing, is con cerned: there fore the best es ti mate plus un cer - tainty must ac count for rules and reg u la tions de rived from the fun da men tal radioprotection prin ci ple which im poses the minimization of the im pact of ra di a tions upon hu mans and the en vi ron ment un der any cir cum stance. In the pres ent pa per, the key el e ments of the ap proach are iden ti fied and char ac ter ized. These shall be seen as the sup port for a con sis tent ap pli ca tion of ther mal-hy drau lics to the de sign and safety of wa ter-cooled nu clear re ac tors
"IAEA SPE-2: RELAP5/MOD2 pre-test calculations and comparison with data measured in the Hungarian PMK-NVH experimental facility"
No full textThe document deals with the description of results obtained by two Relap5/mod2 code versions in the pre-test blind simulation of an experiment performed in the VVER (water cooled- Water Moderated Energy Reactor) experimental simulator PMK-NVH installed at the Budapest Atomic Energy Research Institute in Hungary. The Relap is the well-known computer code developed at Idaho National Laboratory in US: the code is in use at UNIPI since more than a decade. The PMK-NVH loop is an Integral Test Facility (ITF) simulating with reduced height, full pressure, modulated linear power a Russian (Gidropress) type VVER-440. The concerned test was selected by International Atomic Energy Agency (IAEA) as Standard Problem Exercise 2 (SPE 2). The document describes the results of the pre-test calculation submitted (by UNIPI) to Hungary before the execution of the test. This is called blind pre-test analysis: the comparison of about 40 calculated time trends with measured data (once the experiment is performed: this is part of the SPE documentation issued as IAEA report) allows an objective evaluation of the capabilities of the computer code and of the code user team in predicting the scenario of an accident. This is relevant for demonstrating the capabilities in evaluating safety margins of existing NPP, with main reference to VVER (present case)
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