1,720,974 research outputs found
OECD CSNI ISP 21, PIPER-ONE test PO-SB-7: post test analysis performed at Pisa University by RELAP5/MOD2 code, OECD CSNI 2nd Workshop on ISP 21, Calci (I), Apr. 13-14, 1989
No full textFollowing the ISP [see below for acronyms] proposal made by UNIPI to OECD/NEA/CSNI in 1985[proposal given below under quotation] and the first ISP 21 Workshop held in Marina di Grosseto, 1986, the second and final Workshop for the international activity was held in Calci (Pisa geographical region) in 1989.
“The document deals with the proposal made by UNIPI to the OECD/NEA/CSNI (Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Installations) to perform an International Standard Problem (ISP), which was, later on, called ISP 21. This was at the time the first ISP proposed by Italian Institutions dealing with an Integral Test Facility (ITF). ISP 21 was a Small Break Loss of Coolant Accident (SBLOCA) scenario expected to occur in Boiling Water Reactors. One key feature of the proposed experiment was its Counterpart-Test feature which allowed the comparison with similar experiments performed in the FIST facility (available in US, San Jose, General Electric) and the ROSA-III facility (available in Japan at the JAERI research center of Tokai-Mura).
The proposal was accepted and the ISP activity went on in the period 1985-1989. (Later) comparison between experimental scenarios in the three ITF PIPER-ONE, ROSA-III and FIST largely contribute to addressing the scaling issue which was controversial in nuclear thermal-hydraulics.
PIPER-ONE was a Boiling Water Reactor (BWR) simulator installed at the Scalbatraio Laboratory managed by Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN, now DICI) of University of Pisa.”
Four reports are part of this collection.
This report (2 of 4) discusses the calculation results (not blind) related to PIPER-ONE experiment at the basis of ISP 21 performed by UNIPI.
The activity (among the other things) allowed UNIPI understanding of the capabilities of international Institutions in predicting accident scenarios in NPP
FASTRAP: a very fast-running code for the analysis of Special Transients and SB-LOCAs in PWRs
No full textAssessment and Validation of ECART Code in Support to Severe Accident Radionuclide Transport
No full textPresentation of the assessment and validation activities for the ECART Code in support to the studies on radionuclide transport during a severe accident in a LW
Thermal-hydraulic modeling and severe accident radionuclide transport
No full textThe evaluation of radionuclide transport within a nuclear reactor plant and then to the external environment after an accident that involves severe damage to the fuel rods requires an appropriate evaluation of the thermal-hydraulic conditions that influence both the chemical equilibria among the involved species and the radionuclide retention phenomena. The ENEL Code for the Analysis of Radionuclide Transport (ECART) computer program has been developed for the purpose of unifying reactor cooland and containment system analysis and represents the current state of the art of light water reactor severe accident aerosol codes. New aerosol transport models, like physical resuspension and transport under two-phase flow within the reactor coolant system, are included. The code comprises three modules that deal with aerosol transport, chemical equilibria, and thermal hydraulics, respectively. The recently developed thermal-hydraulic module has been applied to the analysis of transients typically addressed by the code to obtain first indications about the adequacy of the adopted models and about the need for further improvements. A thorough assessment is now needed to achieve confidence in the modeling capabilities of the module. The three modules are presently coupled in the integrated ECART code. The obtained code will be further assessed by application to relevant severe accident scenarios
Implementation and Preliminary Verification of the RELAP5/PARCS Code for Pb-Bi Cooled Subcritical Systems
No full textIn the framework of the Italian research program on Accelerator Driven Systems funded by MURST (Italian Ministry of the University, Scientific and Technological Research), the feasibility and operability of a Pb-Bi cooled ADS prototype are currently under investigation. Development and validation of a suitable tool for studying the dynamical behavior of such a system during operational transients and accidental sequences are clearly crucial issues among the R&D needs. This paper deals with the activity performed by ENEA and University of Pisa with the aim to implement and qualify the coupled thermal-hydraulic and neutronic code RELAP5/PARCS, initially developed for LWRs, and adapt it to Pb-Bi cooled subcritical systems. According to this goal, Pb-Bi eutectic thermal and physical properties have been added to RELAP5 libraries while PARCS module has been modified in order to have the possibility of taking into account fast spectrum and time-dependent neutron source driven subcritical systems. A preliminary verification of the modified coupled code RELAP5/PARCS has been carried out analyzing some basic transients. In all these cases, even if the transient is originated by a purely thermal-hydraulic or neutronic event, the interaction between thermal-hydraulic and neutronic phenomena has to be properly considered. Although an exhaustive validation of the modified coupled code has still to be performed by comparing the simulated transients with the results provided by subcritical system dedicated neutronic codes and, restricted to the thermal-hydraulic behavior, with the experimental results, it seems possible to state that the modified coupled code RELAP5/PARCS can acceptably describe the dynamical evolution of a Pb-Bi cooled ADS system
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