Natural Circulation Limits achievable in a PWR

The present paper deals with the Natural Circulation (NC) phenomenon in Pressurized Water nuclear Reactors (PWR). In the first part, data gathered from relevant experiments in PWR simulators are considered. These allowed the establishment of a flow map that has been used for evaluating the NC performance of various reactor concepts. In the second part, a theoretical study has been completed to assess the power removal capability by NC from the core of a PWR having the current geometric configuration. Taking as reference a PWR equipped with U-tubes steam generators, two-phase conditions occur in the core at power levels less than 20% nominal power. Therefore, for core power larger than this value the reactor cannot be classified any more as a PWR. The study shows that from a thermohydraulic point of view, the core can operate at power levels close to the current nominal value without experiencing thermal crisis. Limited consideration has been given to the neutronic design of the core

RELAP5/MOD.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL-34

The present document deals with the Relap5/Mod3.2 analysis of the small break LOCA experiment BL-34 performed in LOBI/MOD2 facility. LOBI/MOD2 was a PWR simulator (Integral Test Facility) installed at JRC (Joint Research Center) in Ispra Establishment (1). Volume scaling and core power scaling factors are 1/712, with respect to the KWU Siemens 1300 MWe (3900 MWt) standard reactor. The experiment is originated by a small break in the cold leg (2" equivalent break area in the plant) without the actuation of the high pressure injection system. Low pressure injection system actuation occurs after core dry-out and accumulators intervention is foreseen when primary pressure falls below 4 MPa. The Relap5 code has been extensively used at University of Pisa; the nodalization of LOBI facility has been qualified through the application of the version Relap5/Mod2 to the same experiment and another test performed in the same facility. Sensitivity analyses have been addressed to the influence of several parameters (like discharge break coefficient, time of accumulators start etc.) upon the predicted transient evolution. Qualitative and quantitative code calculation accuracy evaluation has been performed

US NRC NUREG - RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-03

The present document deals with the Relap5/Mod3.2 analysis of the Small Break LOCA experiment SP-SB-03 performed in SPES facility. SPES is a PWR simulator (Integral Test Facility) installed at SIET center in Piacenza (I). Volume scaling and core power scaling factors are 1/427, with respect to the Westinghouse 900 MWe standard reactor. The experiment is originated by a small break in the cold leg (2" equivalent break area in the plant) without the actuation of the high pressure injection system. Low pressure injection system actuation occurs after core dry-out. The Relap5 code has been extensively used at University of Pisa; the nodalization of SPES facility has been qualified through the application of the version Relap5/Mod2 to the same experiment and another test performed in the same facility. Sensitivity analyses have been addressed to the influence of several parameters (like discharge break coefficient, time of accumulators start etc.) upon the predicted transient evolution. Qualitative and quantitative code calculation accuracy evaluation has been performed

RELAP5/MOD3 .2 Post Test Analysis and Accuracy Quantification of SPES Test SP-SB-04

The present document deals with the Relap5fMod3.2 analysis of the small br.!ak LOCA experiment SP-SB-04 performed in SPES facility. SPES is a PWR simulator (integral Test Facility) installed at SIET center in Pia::enza (IT). Volume scaling and core power scaling factors are 1/427, with respect to the Westinghouse 900 MWe standard reactor. The experiment is originated by a small break in the cold leg (2»> equivalent break area in the plant) without the actuation of the high pressure injection system. The test starts from full power and is the counterpart of the test SP-SB-03, that started at an initial power roughly equal to 10% of nominal power. Low pressure injection system actuation occurs after core dry-out. The Relap5 code has been extensively used at University of Pisa; the nodalizatic'n of SPES facility has, been qualified through the application of the version Relap5/Mod2 to the same experiment and another test performed in the same facility. Sensitivity analyses have been addressed to the influence of several parameters (like discharge break coefficient, time of accumulators start etc.) upon the predicted transient evolution. Qualitative and quantitative code calculation accuracy evaluation has been performed.

USNRC-NUREG - RELAP5/MOD3.2 Post Test Analysis and Accuracy Quantification of Lobi Test BL-44

The present document deals with the Relap5/Mod3.2 analysis of the small break LOCA experiment BL-44 performed in LOBI/MOD2 facility. LOBI(MOD2 was a PWR simulator (Integral Test Facility) installed at JRC (Joint Research Center) in Ispra Establishment (I). Volume scaling and core power scaling factors are 1/712, with respect to the KWU Siemens 1300 MWe (3900 MWt) standard reactor. The expeiment is originated by a small break in the cold leg (2" equivalent break area in the plant) without the actuation of the high pressure injection system. Low pressure injection system actuation occurs after core dry-out and accumulators intervention is foreseen when primary pressure falls below 4 MPa. The Relap5 code has been extensively used at University of Pisa; the nodalization of LOBI facility has been qualified through the application of the version Relap5/Mod2 to the same experiment and another test performed in the same facility. Sensitivity analyses have been addressed to the influence of several parameters (like discharge break coefficient, time of accumulators start etc.) upon the predicted transient evolution. Qualitative and quantitative code calculation accuracy evaluation has been performed

OECD/CSNI ISP 33: Post-test analysis of the PACTEL natural circulation experiment performed by CATHARE 2 v1.3E Code", OECD CSNI Final Workshop on ISP 33 - Lappeenranta (SF), May 17-20, 1993

The document deals with the description of results obtained by the Relap5 code in the post-test simulation of the Natural Circulation (NC) scenario / experiment performed in the Russian Pressurized Water Reactor (WWER = Water cooled, Water Moderated Energy Reactor) experimental simulator PACTEL installed at the Lappenranta University Research Center in Finland. The Relap5 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 PACTEL I a scaled down Integral Test Facility (ITF) simulating with full height, full pressure, full linear power a Russian type 6-loops WWER-440. The concerned test was selected as International Standard Problem 33 (ISP 33) by OECD/NEA/CSNI (Organization for Economic Cooperation and Development / Nuclear Energy Agency / Committee on the Safety of Nuclear Installations). The document describes the results of the post-test calculation submitted (by UNIPI) to Lappeenranta University after the execution of the test. This is called open post-test analysis: the comparison of about 40 calculated time trends with measured data allows an 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 WWER-440 (in this case)

Analysis of single phase natural circulation experiments with Cathare v1.3u and Relap5/mod3.2 codes

During the period 1993 -2004, D’Auria had the opportunity to cooperate with the University of Genova (noticeably Prof. Misale). This cooperation brought to a valuable contribution to the scientific community based on the characterization of the single phase instabilities in a simple atmospheric pressure ‘glass’ loop previously available by UNIGE and upgraded within the framework of the cooperation with UNIPI. The characterization was made with the help of system thermal-hydraulic codes and confirmed some early investigation results by Velander 1968. The present document deals with the application of both RELAP and CATHARE to the analysis of unstable experimental data in the UNIGE loop and was performed in the framework of a cooperation with the Argentinean government institution CNEA (Commission for Nuclear Energy in Buenos Aires)

Application of the FFT Method to the IAEA SPE-2 and SPE-4

The 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 were 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 (Monica Frogheri) has been invited for a few months in Ljubljana at the headquarters of Josef Stefan Institute (JSI) to implement and to apply the method. The present document has been issued by JSI and describes the application of the method to Standard Problem Exercises (SPE) carried out in the framework of the cooperation of JSI and International Atomic Energy Agency (IAEA)

Identification and Categorisation of Safety Issues for ESNII Reactor Concepts. Part I: Common phenomena related to materials

With the aim to develop a joint proposal for a harmonised European methodology for safety assessment of advanced reactors with fast neutron spectrum, SARGEN_IV (Safety Assessment for Reactors of Gen IV) Euratom coordination action project gathered together 22 partners’ safety experts from 12 EU Member States. The group consisted of eight European Technical Safety Organisations involved in the European Technical Safety Organisation Network (ETSON), European Commission’s Joint Research Centre (JRC), system designers, industrial vendors as well as research & development (R&D) organisations. To support the methodology development, key safety features of four fast neutron spectrum reactor concepts considered in Deployment Strategy of the Sustainable Nuclear Energy Technology Platform (SNETP) were reviewed. In particular, outcomes from running European Sustainable Nuclear Industrial Initiative (ESNII) system projects and related Euratom collaborative projects for Sodium-cooled Fast Reactors, Lead-cooled Fact Reactors, Gas-cooled Fast Reactors, and the lead-bismuth eutectic cooled Fast Spectrum Transmutation Experimental Facility were gathered and critically assessed. To allow a consistent build-up of safety architecture for ESNII reactor concepts, the safety issues were further categorised to identify common phenomena related to materials. Outcomes of the present work also provided guidance for identification and prioritisation of further R&D needs respective to the identified safety issues.JRC.F.5-Nuclear Reactor Safety Assessmen