Design Basis Accident Research Articles

Coupled 3D thermal–hydraulic and neutron kinetic Relap5 model for transient analysis of a 1200-MWe nuclear PWR plant

This paper deals with the assessment of the fully integrated three-dimensional thermal–hydraulic and neutron kinetic modeling capability of the Relap5-3D system code. The interest in integrated system codes is growing in the field of safety analyses, where coupled three-dimensional simulations can significantly improve the level of understanding of spatial core phenomena in accident conditions. For this purpose, an asymmetric design basis accident, namely a main steam line break, has been simulated for a commercial 1200-MWe pressurized water reactor. The simulation has been performed with a one-dimensional plant model, including all the major primary, secondary and safety systems included in advanced PWR designs, upgraded with a three-dimensional thermal–hydraulic representation of the reactor vessel and a three-dimensional neutron kinetic core model. The analysis demonstrated the code capability to capture 3D core phenomena, not achievable with 1D plant model based on point kinetic.

Integrity Assessment of RBMK Physical Safety Barriers during Prolonged Station Power-Unit De-Energization

Two successive stages of the control of a beyond-design basis accident with de-energization of an RBMK reactor together with multiple equipment failures are analyzed: depressurization of the circulation loop and restoration of reactor cooldown with cold water as the reactor subsequently warms up. The factors affecting the pressure surge in the circulation loop on restoration of reactor cooling as the reactor heats up are examined. It is shown that the sequence of control actions and the employed equipment are effective from the standpoint of the possibility of overpressurization of the channel tubes and the reactor space. The state of the channel tubes and the reactor space in the presence of arising deformation loads was analyzed using an experimentally determined criterion for the acceptability (integrity) of a channel tube.

Integral LOCA fragmentation test on high-burnup fuel

Increasing the fuel burnup limit in light water reactors is sought to enhance fuel cycle economics and requires establishing a technical basis. Experimental observations of severe fuel fragmentation under loss-of-coolant-accident conditions at Halden and Studsvik had raised the need for additional considerations during the development of this technical basis. These test data suggested that the burnup threshold for high burnup fuel fragmentation may be influenced by pre-transient power. Additional loss-of-coolant-accident test data is therefore valuable to complement these tests and enhance the current state of understating. Oak Ridge National Lab has developed the Severe Accident Test Station capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident and beyond design basis accident scenarios. Severe Accident Test Station provides various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident testing, defueled oxidation quench testing, and high-temperature beyond design basis accident testing. Severe Accident Test Station has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory at Oak Ridge National Lab. Furthermore, descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident test train are provided. Installation of the Severe Accident Test Station system and in-cell re-fabrication restores United States capability to examine postulated and extended loss-of-coolant-accident conditions on spent fuel and cladding and provides a platform for evaluating advanced fuel and accident-tolerant fuel cladding concepts. Lastly, three in-cell integral loss-of-coolant-accident test were performed in the Severe Accident Test Station and compared to the Nuclear Regulatory Commission sponsored loss-of-coolant-accident test as well as the original loss-of-coolant-accident test performed at Argonne National Lab. The results of these tests as well as all publicly available integral loss-of-coolant-accident test were used to validate a threshold for high burnup fuel fragmentation.

Investigation of seismic responses of reactor vessel and internals for beyond-design basis earthquake using elasto-plastic time history analysis

Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake. Elasto-plastic analysis methods are required, and the factors that affect the elasto-plastic behavior of reactor materials should be taken into account. In this study, a material behavior model was developed that considers the irradiation embrittlement effect, which affects the elasto-plastic behavior of the reactor material. This was used to perform the elasto-plastic time history analyses of the reactor vessel and its internals for beyond design basis earthquake. For this investigation, appropriate beyond design basis earthquakes and reliable finite element models were used. Based on the analysis results, consideration was given to the load reduction effect and the margin change. These were transferred to the internals due to the plastic deformation of the reactor vessel.

Design and development of a comprehensive program for the assessment and analysis of environmental effects due to the release of radioactive materials from the stack of nuclear installations: A case study in Bushehr nuclear power plant

The current investigation presents a comprehensive program called KIANA to assess and analyze the environmental effects due to the release of radioactive materials from the stack of nuclear installations. Bushehr Nuclear Power Plant Unit One (BNPP-1) is modeled using the KIANA software and the release of radioactive materials from the BNPP-1 stack on the members of the public through the airborne pathway is evaluated during normal operation as well as in the accident conditions including Design Basis Accidents (DBA) and Beyond Design Basis Accidents (BDBA). To verify and validate the KIANA software, the results of the KIANA are compared with the results of DOZA_M, ESTE, PC-CREAM 98, RECASS Express, and the values of the total effective dose monitored by the local detectors of BNPP-1. The KIANA is developed based on the Gaussian diffusion model written using C# programming language. In the current research, the total effective dose received by a member of the public due to the radioactive plume passage through the airborne pathway is calculated in the normal condition of BNPP-1. Moreover, the total effective dose in the case of the primary-to-secondary leakage inside the steam generators, the total effective dose in the case of Large Break Loss of Coolant Accident (LBLOCA) and Small Break Loss of Coolant Accident (SBLOCA), and the equivalent dose of the thyroid gland tissue for an infant group (1–8 years old) in the case of LBLOCA in DBA conditions are evaluated. Finally, the absorbed dose of the whole body of adults at the initial stage after a BDBA, absorbed dose of the thyroid gland tissue for an infant group (1–8 years old) at the initial stage after a BDBA, and the total effective dose in the first year after the accident in the case of a BDBA are assessed. The results of the KIANA software indicate a good agreement with the results of the DOZA_M, ESTE, PC-CREAM 98, RECASS Express computer programs, and the values of the total effective dose monitored by the local detectors of BNPP-1. The developed software has the potential of calculations of concentration and radionuclide dose received through all the exposure pathways such as airborne, foodstuff, marine, soil, animals, and vegetation without any restriction in normal and accident conditions, simultaneously.

THE VALIDATION OF COMPREHENSIVE SOFTWARE FOR THE ASSESSMENT OF RADIOLOGICAL IMPACTS ON HUMANS THROUGH BENCHMARK ANALYSIS WITH RADIOLOGICAL CODES.

In this paper, a novel domestic code called KIANA was developed for the assessment of radiological impacts on the population in normal and accident conditions including design basis accident (DBA) and beyond DBA (BDBA) for the nuclear power plants. The validation process of the KIANA code was performed using the results of the DOZA_M radiological code, whose results are presented in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant Unit One (BNPP-1). The calculations of KIANA are performed based on the Gaussian diffusion model. The developed KIANA code has the potential of calculating the concentration and radionuclide doses due to the pathways such as airborne, foodstuff, marine (both one and two boxes models), soils, animals, vegetation (with and without tritium) and other pathways without any restriction. In the current research, the individual dose from a cloud to the member of the public in the region of BNPP-1 in normal condition was calculated. Moreover, the total effective dose to the member of the public from the primary to the secondary leakage inside steam generators, large break loss-of-coolant accident (LBLOCA) and small break loss-of-coolant accident was calculated. Thyroid gland equivalent dose for the infant (1-8 years) in the case of LBLOCA at the BNPP in DBA conditions was also evaluated. Finally, the prevented dose at the initial stage for the whole body of adults after BDBA, prevented dose at the initial stage for the thyroid gland of children after BDBA and the effective dose during the first year after the accident (external body irradiation from presence in the area) in the case of BDBA are assessed. The KIANA simulation results showed a good agreement with the FSAR data of BNPP.

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Additionally, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.

Modeling the emergency processes proceeding in the containment of NPP with water-water energy reactor

The article presents the results of computational modeling of the development of accidents in the containment of NPP with water-water energy reactor (WWER) using the COCOSYS software. The containment of the Balakovo NPP with WWER-1000/V-320 was selected as the object of study. The results of the calculation of environmental parameters in the containment during a beyond design basis accident with a “large” leak from the primary circuit at the following initial events: break of the main circulation pipeline DN850 with a full cross section at the reactor inlet, with simultaneous failure of all AC sources including diesel generators for a long period, without operating personnel intervention. A comparison of the results of calculations with the results obtained using the ANGAR software certified by Rostekhnadzor is performed. The calculation results are in good agreement: emergency scenarios are similar; the maximum deviation from absolute pressure values is less than 10 %. Thus, it was concluded that the analysis of processes occurring in the NPP containment during an accident can be performed using the COCOSYS code. This software can be used to independent safety assessment of NPP required by the IAEA standards.

Rapid accident source term estimation (RASTE) for nuclear emergency response in high temperature gas cooled reactor

Rapid accident source term estimation (RASTE) is one of the key functions required by nuclear power plant (NPP) emergency. The results of RASTE are provided to the nuclear decision-making system in emergency, response to ranking accident by INES table and determining suitable protection. This paper proposes the RASTE system for high temperature gas-cooled reactor (HTR) to estimate accident source term based on the simplified methodology used in safety analysis report (SAR) and more realistic operation condition before and in an accident. RASTE simplifies the expression for normal radioactivity in the primary loop, re-suspended radioactivity, additional radioactivity release during the heating process, graphite dust generation, desorbed radioactivity from corrosion of graphite matrix and radioactivity released from failed coated fuel particle. Modifications are made to correct the different ingress water amounts and different fuel burnup. Some unclearly known parameters and processes are estimated from the operation experience as well as accident analysis by conservation rule. The integrated code of fundamental framework and user interface are also developed. This paper takes the High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM), a demonstration project being constructed at Shidao Bay, Shandong Province, China, as the study case to discuss the applicability of RASTE. Considering three typical HTR-PM design basis accidents (DBAs), this paper selects P2, P4 as the representatives of LOCA accidents, selects SG6 and SG91 as the representatives of ingress accidents and selects T2 as the representative of transient accidents. Accuracy verification is made by comparing the environmental release estimated by RASTE to that in SAR. These compared targets include nuclide groups of inert gas, iodine, and aerosols, and key nuclides of Xe-133, I-131, Cs-134. Calculation results show that the methodology presented and RASTE is fit to estimate accident of HTR-PM. The development of RASTE system will contribute to form adequate capability of rapid emergency response for HTR nuclear power plant.

ОСОБЕННОСТИ АКТИВНЫХ ЗОН ПЕРСПЕКТИВНЫХ БЫСТРЫХ РЕАКТОРОВ

In the pre-Chernobyl period, the task of fast reactors was to provide fuel for the intensive development of nuclear energy with a deficit of natural uranium: short doubling time, high HF, high heat intensity, short campaign. After the Chernobyl accident, the main thing was to ensure safety, including in beyond design basis and postulated accidents. Modern fast sodium reactors must, on the one hand, satisfy the safety requirements in design and beyond design basis accidents (in accordance with the requirements of GEN-IV), and on the other, be competitive in comparison with water reactors and other energy sources. The article discusses the methods of transformation of the core of existing and planned Russian fast reactors to meet these requirements. The safety of these reactors in design and beyond design basis accidents is ensured by rejecting the upper end shield and replacing it with a sodium cavity, which allows reducing the NER and ensuring the input of negative reactivity when sodium boils. Such an core is implemented in BN-800 and the designed BN-1200. Improving the competitiveness (increasing the duration of the fuel campaign and KIUM) of the designed and existing reactors is ensured by switching to radiation-resistant clad steel EK164 (currently sold in BN-600, planned in BN-800) and the transition to the core with an axial layer of depleted uranium dioxide (planned for BN-800 and BN-1200). In BN-1200, due to the abandonment of a number of switchgear elements and optimization of the design, a one-and-a-half reduction in metal consumption is achieved. For fuel economy, a ″thick″ fuel rod with a diameter of 9.3 mm and a fuel campaign of 4 years have been adopted. This reduces the annual consumption of fuel rods and fuel assemblies by 2-2.5 times. Both events require modernization of the core. Each of these events allows you to increase the duration of the campaign by 25 %, and the implementation of both measures (transition to new steel and the introduction of a layer) will allow you to increase the campaign and reduce fuel consumption by more than 50 %. The introduction of the axial layer does not require core expansion, which gives additional fuel savings. But the axial layer should not be located in the central plane of the core, but with a shift of 3-5 cm downward. This will improve many characteristics of the core: increase fuel enrichment by a quarter and, as a result, proportionally reduce the neutron flux and radiation damage to the claddings of fuel elements; increase the fuel campaign to 5 years; to reduce the reactivity margin for burnout, without compromising the effectiveness of the control systems, to increase the micro-campaign to 1 year; to level the heat-release field in the vertical direction, to reduce the maximum heat intensity; to improve the course of beyond design basis accidents like ULOF. This method of reducing the rate of accumulation of a damaging dose is equally effective for both MOX and nitride fuels, which makes it possible to unify the core design.

System of Laser Monitoring of Water Pollution with Application of Relative Description of Signal Shape

As a rule, the wastewater treatment system is not designed to filter substances formed, as a result of beyond design basis accident. The nature of the beyond design basis accident is associated with the shortterm appearance of a clot of these substances in wastewater, determined by the volume of the substance storage tank. Therefore, a rational approach is to divert this portion of the formed substances into a separate branch of the sewage system or sedimentation tanks. The aim of the work is to implement this approach by creating a laser monitoring system for water pollution.The article proposes a system for automatic detection of a clot of emergency discharge of pollutants into the wastewater of an industrial enterprise. The structural diagram of the system and the purpose of its main elements are given. The system should provide clot detection in real time. To ensure this function, a preliminary study is made of the spectral characteristics of all substances that may appear in wastewater in the event of an emergency.Based on these data, the wavelengths of laser radiation in the system are selected. The obtained measurement data from several probes are presented in the form of a lattice function, which is translated into a relative description representing the order relationship matrix on the set of lattice function components. The relative description is invariant to linear changes in the lattice function. The decision to detect any substance from emergency discharges is made based on a comparison of the relative description of the measurements with the standards prepared at the stage of system setup.The article provides an example of the formation of standards for emergency clots from glycerin and allyl alcohol. The graphs of the lattice functions obtained from the IR spectra of emergency discharges of these substances are given; algorithms for constructing a lattice function and comparison of lattice functions. Thus, using the developed mathematical description of the shape of digital signals based on the relative description, the signal of the monitoring curve can be described in the form of a curve of the optical density change of an aqueous medium.

УЧЕТ НЕОПРЕДЕЛЕННОСТЕЙ В ИСХОДНЫХ ДАННЫХ ПРИ ПРОВЕДЕНИИ РАСЧЕТОВ ЭФФЕКТИВНОГО КОЭФФИЦИЕНТА РАЗМНОЖЕНИЯ АКТИВНОЙ ЗОНЫ ИССЛЕДОВАТЕЛЬСКОЙ ЯДЕРНОЙ УСТАНОВКИ

Estimation of the error in calculations, that is caused by uncertainties in the initial data is an integral part of the nuclear facilities safety case. Nuclear safety is one of the fundamental directions in safety case. Incorrect accounting for the uncertainties in the initial data can lead to a row of negative consequences, in particular, incorrect determination of the reactivity margin in various modes of normal operation, as well as in cases of the abnormal operation of nuclear facilities. This article presents results of a research of uncertainties influence on an effective multiplication factor of research reactor PIK. The design basis accident (deliquation of the central experimental channel) was analyzed during the discussed research.

Innovative passive management of the Station Black-Out accident for VVER1000/V446 NPP and development of its emergency operating procedures

Loss Of Offsite Power (LOOP) accident is one of the most important Design Basis Accidents (DBAs) that originate due to loss of the house-load demands which make serious concerns for each NPP. Moreover, in the case of loss of all diesel generators (due to earthquake, etc.), the accident forwards to the Station Black-Out(SBO). Following and learning the Fukushima accident, the cares have been focused on using internal and/or external water resources for the passive management of the SBO accident. In the current study, passive strategies in the VVER-1000/V446 NPP are proposed using a tank in the secondary loop (called Deaerator) for the SBO management. After RELAP5 modeling benchmarking in the steady-state as well as standard SBO management (which are mentioned in the NPP FSAR), two operational SBO management are proposed herein including “Feeds and Bleeds” (in the primary and secondary loops) and also including a Deaerator tank in the secondary side. The results show that using the Deaerator-water tank of the secondary side as well as feed and bleed, the time available for operators before the core heat-up (i.e. fuel cladding temperature of 1200 °C); extends to 35000 s. Obviously; it gives a good opportunity for plant operators to access further time for using other facilities/methods to avoid core damage.

Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions

A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAl cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAl alloy cladding and Zircaloy-2 cladding were subjected to rapid 300°-650 °C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 °C. In the range of 32–55 MPa hoop stress, Zircaloy-2 cladding burst within 1–16 cycles (about 100 s of dryout duration above 600 °C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 °C). Higher temperature 300°-700 °C and 300°C–800 °C cycling experiments had to be performed to induce C26 M burst – failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2.

Plastic damage analysis of pile foundation of nuclear power plants under beyond-design basis earthquake excitation

With the rapid development of nuclear industry in China, an increasing number of nuclear power plants (NPPs) are unavoidable to be constructed upon non-lithological foundations. Therefore, it is essential to optimize the current evaluation methods for non-lithological foundation reinforcement plans. By using the Octree discretization method combining with the scaled boundary polyhedron finite element method (SBFEM), a refined calculation model of a Generation III+ NPP is constructed. Due to the highly nonlinear dynamic characteristics of the system involved, this paper adopts a generalized plasticity model for the soil, a plastic damage model for the piles, and a nonlinear overburden layer-based seismic input method. Subsequently, large scale plastic damage analysis in detail to the NPPs under beyond-design basis earthquake inputs is performed. The results show that the pile foundation damage characteristics and distribution acquired from simulations agree well with the data from actual earthquake disasters. The most severe damage to the pile foundation occurred at the interface between the end of the pile and the base mat. The inclusion of plastic damage showed that the floor response spectra of the nuclear island experiences significant changes in three dimensions. Therefore, the effect of pile damage to the dynamic response of the power plants should be considered for strong earthquake excitation. The computational and analysis methods proposed in this paper can effectively reflect the dynamic nonlinear response of the nuclear power plants pile foundation-soil interaction.

SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept

Abstract In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Safety Analysis of the TRIGA 2000 U3Si2-Al Fuel Core Under Reactivity Insertion Accidents

TheTRIGA 2000 reactor in Bandung is planned to change its fuel type from the TRIGA fuel rod type to the U 3 Si 2 -Al plate type of low enriched uranium of 19.75 % with uranium density of 2.96 gU/cc. A study on the neutronic parameters from the equilibrium core has been done. To ensure safe operation of the new fuel, thermodynamic evaluation of the core needs to be done. The purpose of this study is to conduct a reactor safety analysis of reactivity insertion during withdrawal of the control rod and to study the effect of this reactivity insertion on the power and the maximum temperature of the fuel and the cladding. Reactivity insertion accident is the main factor of the design basis accidents in nuclear reactor design. A simulation of transient for reactivity insertion has been carried out using a coupled neutronic and thermal-hydraulic MTR-DYN code. The code was developed based on three-dimensional multigroup neutron diffusion theory. The coupled space and time-dependent problem were solved by adiabatic model. Transient analysis was performed for a reactivity insertion of 32.33 pcm/s with the assumption that all of the control rods were rapidly withdrawn. For the insertion at a low power of 100 W, the maximum power achieved was 2.74 MW while a maximum power of 2.3 MW was achieved for the power transient of 1 MW. The maximum temperature of the coolant, the cladding, and the fuel for TRIGA 2000 core does not exceed the allowable safety limit for reactivity insertions . Normal 0 false false false EN-US X-NONE X-NONE /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Table Normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:10.0pt; font-family:"Times New Roman","serif"; mso-ansi-language:EN-US; mso-fareast-language:EN-US;}

A summary of fission-product-transport phenomena during SGTR severe accidents

A steam-generator-tube rupture (SGTR) in a pressurized-water reactor is a design-basis accident with which plants have been designed to cope. Its specific significance is in its potential to bypass the containment thereby providing a pathway for radioactivity release to the environment. In combination with significant core damage, uncontrolled SGTR may lead to accidental release of radioactivity in the form of aerosol particles and gaseous compounds.In this paper, we describe the phenomena identification and ranking tables (PIRT) prepared as an international collaboration for uncontrolled SGTR in connection with core damage. This PIRT is limited to fission-product transport and retention in the steam generator and release to the auxiliary building, i.e., the source term. The phenomena affecting the fission-product retention in the steam generator and release to the environment were divided into those relevant in a dry steam-generator secondary side, and those relevant in a flooded secondary side. Rankings were given to 17 phenomena in the dry and 16 phenomena in the flooded secondary side regarding their: i) impact on the source term, ii) the availability of experimental data applicable to SGTR severe-accident conditions, and iii) the availability of codes/models to describe the phenomenon under the conditions of SGTR severe accidents.The SGTR boundary conditions and the considered phenomena are briefly described. The results are given in the form of ranking tables separately for the dry and flooded steam generator secondary side. The most important phenomena and their interactions with other phenomena are discussed with references to existing data and models. The most important uncertain phenomena in the dry steam generator scenario are particle bounce and fragmentation; whereas inertial impaction and retention by droplets in the jet region are the most uncertain ones in the flooded configuration. The need for improved knowledge is, though, tempered by overall high aerosol retention in the latter scenario. The applicability of the existing data and models for SGTR accident analysis is considered in the discussion.

Total and partial loss of coolant experiments in an instrumented fuel assembly of IEA-R1 research reactor

The safety of nuclear facilities has been a growing global concern, mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), many times considered a design basis accident, are important for ensure the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and it is necessary to assure the decay heat removal as a safety condition. This work aimed to perform, in a safe way, partial and complete uncovering experiments for an Instrumented Fuel Assembly (IFA), in order to measure and compare the actual fuel temperatures behavior for LOCA in similar conditions to research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 core and positioned in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. Experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. It was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases, for the specific conditions of heat decay intensity and dissipation analyzed. The maximum temperatures reached in all experiments were quite below the fuel blister temperature, which is around 500 °C. The STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.