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Mirea Mladin

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Papers by Mirea Mladin

RELAP5 Calculations for Loss of Cooling Test at Romanian TRIGA 14 MW Research Reactor

During Romanian TRIGA 14 MW steady state reactor commissioning in 1979 and beginning of 1980 a se... more During Romanian TRIGA 14 MW steady state reactor commissioning in 1979 and beginning of 1980 a series of tests were performed to check the functioning of reactor systems and the response of the reactor core to transient initiators. Two of these tests consisted in inducing total forced cooling loss and also loss of primary pumps cooling with emergency pump in operation for 15 minutes after the event. Maximum fuel temperatures in hot pins were recorded. A RELAP5 reactor model was created to simulate the Loss of Coolant Transients and to compare with experiment. The scope of the paper is to describe the model and the most significant parameters which affect the results and to compare RELAP5 fuel temperature with that observed during the experiment in the hottest TRIGA fuel element.

Evaluation of the RELAP5/SCDAP Accident Analysis Code Applicability to Candu Nuclear Reactors

Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor ... more Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor de tip LWR, fiind rezultatul cuplării dintre RELAP5 pentru analiza termohidraulică şi SCDAP pentru modelarea accidentelor severe. Lucrarea trece în revistă rezultatele aplicării codului RELAP5 la reactorii de tip CANDU şi prezintă rezultatele obţinute pentru analiza fazei iniţiale a unui accident sever pentru un reactor de tip CANDU. SCDAP/RELAP5 code has been developed in US for best-estimate simulation of light water reactors transients during nuclear accidents. It is the result of the coupling between RELAP5, modelling thermalhydraulic, and SCDAP, modelling the behavior of the reactor core during severe accidents. RELAP5 applications to CANDU will be reviewed in this paper in the frame of evaluation of the thermalhydraulic analysis capabilities of the code. Particular SCDAP early phase models as well as late phase models covering a wide range of phenomena can be applied to CANDU, as proven in the paper.

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

Nuclear Engineering and Design, Oct 1, 2009

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the de... more The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4-CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4-CANDU code predictions.

Simulation of B9401 Test in the RD-14M Experimental Facility With CATHARE2

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 2010

ABSTRACT RD-14M experimental facility is a full vertical-scale representation of a CANDU heat tra... more ABSTRACT RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.

Investigation on Header Manifold Model Effect at CANDU6 Large Loss of Cooling Accident

Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance, 2008

ABSTRACT In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pre... more ABSTRACT In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pressure tubes tube through which the heavy-water coolant flows. 95 pressure tubes are connected by individual feeders to a common header. For CANDU 6 safety analyses, even when multiple channels model is employed, only one node is used for header. In this approach, all the channels are subjected to the same boundary condition. However, site inlet and outlet header pressure measurements and ultrasonic feeder flow data, confirm the existence of axial pressure gradients along the inlet and outlet headers. These axial pressure gradients would give rise to individual header-to-header pressure drops for each channel and also to flow distribution throughout both the inlet and outlet headers. In this paper, the header manifold model effect on the large break loss of coolant accident analyses of CANDU reactors has been performed by RELAP5/ mod 3.4 code. The 35% reactor inlet header break was selected for this study, as this break size produce the highest fuel clad temperature among all postulated breaks size. The results obtained considering the header manifold model, show that location of fuel channel upon break location has a strong impact on peak clad temperature calculation.

Verification of Modified SCDAP/RELAP5 Against CANDU Fuel Deformation Experimental Results

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications, 2009

ABSTRACT Two new models for treating the fuel channel during the early phase of a severe accident... more ABSTRACT Two new models for treating the fuel channel during the early phase of a severe accident in CANDU reactors were recently created and tested in SCDAP/RELAPSIM/MOD3.4. These treat the deformations phenomena and the possible relocation of metallic melt mixture inside the fuel channel, respectively. The horizontal relocation model consists in a series of modifications in the original Relocation-Flow-Solidification (LIQSOL) model in SCDAP code. The purpose of the paper is to apply the code to an experiment dedicated to CANDU fuel bundle behavior during degraded cooling and to attempt a validation of the model. The work generates and discusses the results of the modified SCDAP/ RELAP5 on a particular experiment in which a 37-element CANDU bundle was subjected to superheated steam inside a ballooned pressure tube. The focus is on the temperature evolution, the hydrogen accumulation and the metallic melt final relocation geometry.

SCDAP/RELAP5 Analysis of Progression of Severe Accident in CANDU 6 Plant

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance, 2009

ABSTRACT The CANDU system design has specific features which are important to severe accidents pr... more ABSTRACT The CANDU system design has specific features which are important to severe accidents progression and require selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in CANDU system are largely less validated and, as a consequence, the level of uncertainty remains high in many instances. Unlike the light water reactors, for which several different computer codes to analyze severe accidents exist, for CANDU severe accidents analysis only two codes were developed: MAAP4-CANDU and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented CANDU 6 specific models. Thus, the two codes have many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose of the project is the assessment and adaptation of the SCDAP/RELAP5 code to CANDU 6 severe accidents analysis. The present work investigates the progression of a severe accident in CANDU 6 reactor starting from a LBLOCA and a subsequent loss of all heat sinks. The paper provides details concerning the methodology and nodalisation used, and interprets the results obtained. Comparisons of the SCDAP/RELAP5 simulations with the MAAP4-CANDU and the ISAAC codes reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAP/RELAP5, MAAP4-CANDU and ISAAC codes predictions.

IAEA-SM-360/27P Safe Operation of a Triga Reactor in the Situation of Leu–Heu Core Conversion

$Research paper thumbnail of A prompt gamma neutron activation analysis facility using a diffracted beam$

A prompt gamma neutron activation analysis facility using a diffracted beam

Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 1993

A prompt gamma neutron activation analysis facility has been constructed at the MIT Research Reac... more A prompt gamma neutron activation analysis facility has been constructed at the MIT Research Reactor using a diffracted beam from a multilayered graphite monochromator. A beam of 0.0143 eV neutrons of intensity 6X lo6 n/cm% is available at the sample position. Backgrounds are low due to the use of the diffracted beam and are further improved by a sapphire crystal in the beam line. This design allowed the Ge detecting crystal to be placed close to the sample position, -4 cm, with a resultant high detection efficiency. The sensitivity of the facility is reported for several representative pure elements. The major impetus for the construction of this facility was the need for accurate analyses of l"B in biological samples for neutron capture therapy research. Detailed results for this type of analysis are provided. The sensitivity of this diffracted beam facility currently exceeds that of two representative direct beam facilities using reactors of twice the power of the MITR-II. Possible major improvements in sensitivity, more than an order of magnitude, and in background levels are outlined for future development.

Calculation of radioactive species transport in a TRIGA reactor

Nuclear Engineering and Design, 2013

h i g h l i g h t s

Evaluation of UV Optical Fibers Behavior under Neutron Irradiation

Advances in Materials Physics and Chemistry, 2012

Degradation of UV transmitting optical fibers under nuclear reactor neutron exposure is reported.... more Degradation of UV transmitting optical fibers under nuclear reactor neutron exposure is reported. Four type of optical fibers (solarization resistant, H 2 -loaded; UV transmission standard OH; UV enhanced transmission, high OH, H 2 -loaded; high OH, deep UV enhanced) were exposed to neutron fluences up to 4 x 10 17 n/cm 2 . The optical transmission was measured off-line over the 200 nm -900 nm spectral range and the build-up of color centers was monitored.

Evaluation of the RELAP5/SCDAP Accident Analysis Code Applicability to Candu Nuclear Reactors

Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor ... more Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor de tip LWR, fiind rezultatul cuplării dintre RELAP5 pentru analiza termohidraulică şi SCDAP pentru modelarea accidentelor severe. Lucrarea trece în revistă rezultatele aplicării codului RELAP5 la reactorii de tip CANDU şi prezintă rezultatele obţinute pentru analiza fazei iniţiale a unui accident sever pentru un reactor de

Progress on ruthenium release and transport under air ingress conditions

Nuclear Engineering and Design, 2008

Multiple aspects of raw data collection and processing for Romanian TRIGA SSR

Nuclear Engineering and Design, 2010

Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nucl... more Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nuclear reactor. In general, PSA is used to support the system design, configuration decisions and the operational safety management of the plant. Ideally, failure data used for safety and reliability analyses should be based on site-specific data. The paper presents activities and results that started in 2001 with the participation of Romanian TRIGA Steady State 14 MW reactor in an international project for expanding component reliability database for Research Reactors for PSA use. Collection of reliability data for Romanian TRIGA is discussed emphasizing the problems encountered. Failure data for some components is given, both standalone and in comparison with other research reactors, as well as statistics pertaining to the failures and failure modes of the investigated components. A brief presentation of a software application developed in the Institute for Nuclear Research (INR) for raw data acquisition and processing is included.

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

Nuclear Engineering and Design, 2009

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the de... more The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4-CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4-CANDU code predictions.

SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

Nuclear Engineering and Design, 2009

Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been... more Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR-BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

Fuel behavior comparison for a research reactor

Journal of Nuclear Materials, 2006

The paper presents the behavior and properties analysis of the low enriched uranium fuel, which w... more The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies

Annals of Nuclear Energy, 2010

In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat s... more In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the longterm coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris. (M. Mladin). 1 Dimensions and features for CANDU are generic.

Modifications in SCDAP code for early phase degradation in a CANDU fuel channel

Annals of Nuclear Energy, 2009

The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (... more The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (CANDU) reactor type constitutes the major feature which prevents the direct application to this reactor design of the severe accident physical models developed for PWR/BWRs. During a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency core cooling (LOECC), particular deformation phenomena take place inside a CANDU fuel channel. Another peculiarity appears at melt relocation in horizontal geometry. Brief discussion of accident phenomena is included.

RELAP5 Calculations for Loss of Cooling Test at Romanian TRIGA 14 MW Research Reactor

During Romanian TRIGA 14 MW steady state reactor commissioning in 1979 and beginning of 1980 a se... more During Romanian TRIGA 14 MW steady state reactor commissioning in 1979 and beginning of 1980 a series of tests were performed to check the functioning of reactor systems and the response of the reactor core to transient initiators. Two of these tests consisted in inducing total forced cooling loss and also loss of primary pumps cooling with emergency pump in operation for 15 minutes after the event. Maximum fuel temperatures in hot pins were recorded. A RELAP5 reactor model was created to simulate the Loss of Coolant Transients and to compare with experiment. The scope of the paper is to describe the model and the most significant parameters which affect the results and to compare RELAP5 fuel temperature with that observed during the experiment in the hottest TRIGA fuel element.

Evaluation of the RELAP5/SCDAP Accident Analysis Code Applicability to Candu Nuclear Reactors

Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor ... more Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor de tip LWR, fiind rezultatul cuplării dintre RELAP5 pentru analiza termohidraulică şi SCDAP pentru modelarea accidentelor severe. Lucrarea trece în revistă rezultatele aplicării codului RELAP5 la reactorii de tip CANDU şi prezintă rezultatele obţinute pentru analiza fazei iniţiale a unui accident sever pentru un reactor de tip CANDU. SCDAP/RELAP5 code has been developed in US for best-estimate simulation of light water reactors transients during nuclear accidents. It is the result of the coupling between RELAP5, modelling thermalhydraulic, and SCDAP, modelling the behavior of the reactor core during severe accidents. RELAP5 applications to CANDU will be reviewed in this paper in the frame of evaluation of the thermalhydraulic analysis capabilities of the code. Particular SCDAP early phase models as well as late phase models covering a wide range of phenomena can be applied to CANDU, as proven in the paper.

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

Nuclear Engineering and Design, Oct 1, 2009

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the de... more The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4-CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4-CANDU code predictions.

Simulation of B9401 Test in the RD-14M Experimental Facility With CATHARE2

18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 2010

ABSTRACT RD-14M experimental facility is a full vertical-scale representation of a CANDU heat tra... more ABSTRACT RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.

Investigation on Header Manifold Model Effect at CANDU6 Large Loss of Cooling Accident

Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance, 2008

ABSTRACT In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pre... more ABSTRACT In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pressure tubes tube through which the heavy-water coolant flows. 95 pressure tubes are connected by individual feeders to a common header. For CANDU 6 safety analyses, even when multiple channels model is employed, only one node is used for header. In this approach, all the channels are subjected to the same boundary condition. However, site inlet and outlet header pressure measurements and ultrasonic feeder flow data, confirm the existence of axial pressure gradients along the inlet and outlet headers. These axial pressure gradients would give rise to individual header-to-header pressure drops for each channel and also to flow distribution throughout both the inlet and outlet headers. In this paper, the header manifold model effect on the large break loss of coolant accident analyses of CANDU reactors has been performed by RELAP5/ mod 3.4 code. The 35% reactor inlet header break was selected for this study, as this break size produce the highest fuel clad temperature among all postulated breaks size. The results obtained considering the header manifold model, show that location of fuel channel upon break location has a strong impact on peak clad temperature calculation.

Verification of Modified SCDAP/RELAP5 Against CANDU Fuel Deformation Experimental Results

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications, 2009

ABSTRACT Two new models for treating the fuel channel during the early phase of a severe accident... more ABSTRACT Two new models for treating the fuel channel during the early phase of a severe accident in CANDU reactors were recently created and tested in SCDAP/RELAPSIM/MOD3.4. These treat the deformations phenomena and the possible relocation of metallic melt mixture inside the fuel channel, respectively. The horizontal relocation model consists in a series of modifications in the original Relocation-Flow-Solidification (LIQSOL) model in SCDAP code. The purpose of the paper is to apply the code to an experiment dedicated to CANDU fuel bundle behavior during degraded cooling and to attempt a validation of the model. The work generates and discusses the results of the modified SCDAP/ RELAP5 on a particular experiment in which a 37-element CANDU bundle was subjected to superheated steam inside a ballooned pressure tube. The focus is on the temperature evolution, the hydrogen accumulation and the metallic melt final relocation geometry.

SCDAP/RELAP5 Analysis of Progression of Severe Accident in CANDU 6 Plant

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance, 2009

ABSTRACT The CANDU system design has specific features which are important to severe accidents pr... more ABSTRACT The CANDU system design has specific features which are important to severe accidents progression and require selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in CANDU system are largely less validated and, as a consequence, the level of uncertainty remains high in many instances. Unlike the light water reactors, for which several different computer codes to analyze severe accidents exist, for CANDU severe accidents analysis only two codes were developed: MAAP4-CANDU and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented CANDU 6 specific models. Thus, the two codes have many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose of the project is the assessment and adaptation of the SCDAP/RELAP5 code to CANDU 6 severe accidents analysis. The present work investigates the progression of a severe accident in CANDU 6 reactor starting from a LBLOCA and a subsequent loss of all heat sinks. The paper provides details concerning the methodology and nodalisation used, and interprets the results obtained. Comparisons of the SCDAP/RELAP5 simulations with the MAAP4-CANDU and the ISAAC codes reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAP/RELAP5, MAAP4-CANDU and ISAAC codes predictions.

IAEA-SM-360/27P Safe Operation of a Triga Reactor in the Situation of Leu–Heu Core Conversion

$Research paper thumbnail of A prompt gamma neutron activation analysis facility using a diffracted beam$

A prompt gamma neutron activation analysis facility using a diffracted beam

Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 1993

A prompt gamma neutron activation analysis facility has been constructed at the MIT Research Reac... more A prompt gamma neutron activation analysis facility has been constructed at the MIT Research Reactor using a diffracted beam from a multilayered graphite monochromator. A beam of 0.0143 eV neutrons of intensity 6X lo6 n/cm% is available at the sample position. Backgrounds are low due to the use of the diffracted beam and are further improved by a sapphire crystal in the beam line. This design allowed the Ge detecting crystal to be placed close to the sample position, -4 cm, with a resultant high detection efficiency. The sensitivity of the facility is reported for several representative pure elements. The major impetus for the construction of this facility was the need for accurate analyses of l"B in biological samples for neutron capture therapy research. Detailed results for this type of analysis are provided. The sensitivity of this diffracted beam facility currently exceeds that of two representative direct beam facilities using reactors of twice the power of the MITR-II. Possible major improvements in sensitivity, more than an order of magnitude, and in background levels are outlined for future development.

Calculation of radioactive species transport in a TRIGA reactor

Nuclear Engineering and Design, 2013

h i g h l i g h t s

Evaluation of UV Optical Fibers Behavior under Neutron Irradiation

Advances in Materials Physics and Chemistry, 2012

Degradation of UV transmitting optical fibers under nuclear reactor neutron exposure is reported.... more Degradation of UV transmitting optical fibers under nuclear reactor neutron exposure is reported. Four type of optical fibers (solarization resistant, H 2 -loaded; UV transmission standard OH; UV enhanced transmission, high OH, H 2 -loaded; high OH, deep UV enhanced) were exposed to neutron fluences up to 4 x 10 17 n/cm 2 . The optical transmission was measured off-line over the 200 nm -900 nm spectral range and the build-up of color centers was monitored.

Evaluation of the RELAP5/SCDAP Accident Analysis Code Applicability to Candu Nuclear Reactors

Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor ... more Codul de calcul SCDAP/RELAP5 a fost dezvoltat în US pentru analiza termohidraulică a reactorilor de tip LWR, fiind rezultatul cuplării dintre RELAP5 pentru analiza termohidraulică şi SCDAP pentru modelarea accidentelor severe. Lucrarea trece în revistă rezultatele aplicării codului RELAP5 la reactorii de tip CANDU şi prezintă rezultatele obţinute pentru analiza fazei iniţiale a unui accident sever pentru un reactor de

Progress on ruthenium release and transport under air ingress conditions

Nuclear Engineering and Design, 2008

Multiple aspects of raw data collection and processing for Romanian TRIGA SSR

Nuclear Engineering and Design, 2010

Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nucl... more Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nuclear reactor. In general, PSA is used to support the system design, configuration decisions and the operational safety management of the plant. Ideally, failure data used for safety and reliability analyses should be based on site-specific data. The paper presents activities and results that started in 2001 with the participation of Romanian TRIGA Steady State 14 MW reactor in an international project for expanding component reliability database for Research Reactors for PSA use. Collection of reliability data for Romanian TRIGA is discussed emphasizing the problems encountered. Failure data for some components is given, both standalone and in comparison with other research reactors, as well as statistics pertaining to the failures and failure modes of the investigated components. A brief presentation of a software application developed in the Institute for Nuclear Research (INR) for raw data acquisition and processing is included.

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

Nuclear Engineering and Design, 2009

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the de... more The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4-CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4-CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4-CANDU code predictions.

SCDAP/RELAP5 application to CANDU6 fuel channel analysis under postulated LLOCA/LOECC conditions

Nuclear Engineering and Design, 2009

Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been... more Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR-BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions.

Fuel behavior comparison for a research reactor

Journal of Nuclear Materials, 2006

The paper presents the behavior and properties analysis of the low enriched uranium fuel, which w... more The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies

Annals of Nuclear Energy, 2010

In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat s... more In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the longterm coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris. (M. Mladin). 1 Dimensions and features for CANDU are generic.

Modifications in SCDAP code for early phase degradation in a CANDU fuel channel

Annals of Nuclear Energy, 2009

The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (... more The existence of horizontal fuel channels surrounded by moderator in a CANada Deuterium Uranium (CANDU) reactor type constitutes the major feature which prevents the direct application to this reactor design of the severe accident physical models developed for PWR/BWRs. During a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency core cooling (LOECC), particular deformation phenomena take place inside a CANDU fuel channel. Another peculiarity appears at melt relocation in horizontal geometry. Brief discussion of accident phenomena is included.