Anticipated Transient Without Scram Research Articles

TRACE assessment of density wave instability onset with void reactivity feedback under natural circulation

Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

The Japan Atomic Energy Agency (JAEA) has been developing a plant dynamics analysis code named Super-COPD for design studies and safety analyses of sodium-cooled fast reactors. Since there are a few validation studies of a thermal–hydraulic model in a plant, including a point kinetics model considering reactivity feedback for anticipated transient without scram (ATWS) events classified into the beyond design basis events (BDBEs), further validation studies under BDBE conditions are needed to improve the prediction accuracy of the model. Therefore, JAEA has participated in the international benchmark analysis of loss of flow without scram (LOFWOS) test #13 performed at the Fast Flux Test Facility (FFTF) as one of passive safety demonstration tests hosted by the International Atomic Energy Agency. In the first step, the transient calculation with major reactivity feedback mechanisms and imposing measured power history was conducted with Super-COPD. The results were compared with the measured data. Several challenges were identified about the thermal–hydraulic model in the core and heat transport systems and the reactivity model in the point kinetics model. Consequently, the transient calculation was performed with modified models considering the radial heat transfer, flow redistribution, and interwrapper flow in the core. The time-dependent change and the second peak of the outlet temperature in the specified assembly were predicted accurately because the core thermal–hydraulics was reproduced well. Furthermore, the assembly bowing reactivity model was developed to improve the net reactivity, thereby resulting in a good agreement with the measured data on the reactor power, temperature, and flow rate in the primary loop. Thus, this paper proves the validation of Super-COPD for the LOFWOS event in the BDBE.

Anticipated transients without scram in an innovative power nuclear reactor of anomalously high power

The article considers the possibility of creating a safe power reactor of anomalously high power. This may be relevant in the transition to large-scale safe and environmentally acceptable nuclear power. The safety justification is carried out based on mathematical modeling of the most dangerous emergency modes from among ATWS (anticipated transients without scram). With the use of a number of innovations (the UN-PuN-U cermets fuel used, a coolant based on lead with a high concentration of the 208Pb isotope, and tungsten fuel cladding), it was possible to ensure the safety of a fast reactor with an electric power of up to 3 GW. In contrast to previously proposed innovations, the article considers core layouts containing fuel pins with claddings made entirely of tungsten. Emergency modes are modeled approximately based on solving three groups of equations: a system of point kinetic equations, a non-stationary heat conduction equation, which, and a reactivity balance equation. The value of the void reactivity effect is estimated based on the solution of the gas-kinetic neutron transport equation. Reserves have been identified in further increasing the power and safety of the power reactor. They are associated with the optimization of the size of granules of ceramic and metal powders in the fuel composition.

Preliminary analysis focusing on in-vessel thermal hydraulics in loss-of-heat removal systems in a sodium-cooled fast reactor

Accident scenarios for the safety evaluation of sodium-cooled fast reactors (SFRs) are categorized into two groups: Anticipated Transient Without Scram (ATWS) and Loss-of-Heat Removal System (LOHRS). The latter is subdivided into two scenarios—Loss-of-Heat Sink (LOHS) and Loss-of-Reactor Level (LORL)—depending on the initiating event. With the development of SFRs, research on ATWS has been paid much attention to due to its rapid accident progression, which could lead to significant radioactive release. Conversely, the significance of LOHRS has increased in recent decades due to lessons learned from accidents at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company. However, compared with ATWS, finding and evaluation tools for LOHRS have been limited. Hence, code development is initiated to enhance the severe accident analysis code AZORES, which the Nuclear Regulation Authority developed for the safety evaluation of SFR accidents. This enhancement is achieved by coupling AZORES with plant dynamics code ADYTUM in order to deal with more detailed thermal hydraulics behavior in reactor cooling system. The paper describes the coupling methodology, focusing on the thermal–hydraulic model as the first step of code development and its application. The coupling has been confirmed to be well achieved through a preliminary analysis of LOHS, comparing the results of coupled and single AZORES. The comparison of the analysis results between LOHS and LORL by the AZORES coupled with ADYTUM has been conducted to identify the key phenomena in accident progression. The comparison shows that the coolant leak area and the pressure drop at the piping of the reactor cooling system affect accident progression in LOHS. Consequently, the coupled AZORES is capable of predicting in-vessel thermal–hydraulic behavior in LOHRS successfully, reflecting accident conditions and plant design appropriately.

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. We assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

The concept of a fast reactor with liquid metal fuel in tungsten capsules

The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency modes from among ATWS (anticipated transients without scram) is carried out. The reactor is reliable and safe. The maximum temperature of a high-temperature reactor coolant is close to the boiling point of lead. By the nature of the change in the maximum temperatures of the core components, the reactor occupies an intermediate position between a reactor with solid metallic fuel and a reactor with cermets based on UN-PuN and metal uranium nanopowder.

A study of the behavior of the floating absorber for safety at transient (FAST) in an innovative sodium-cooled fast reactor (iSFR)

Negative reactivity feedback is one of the most important inherent safety characteristics of reactors for suppressing changes in thermal power and component temperatures during accidents. The floating absorber for safety at transient (FAST) has been suggested to improve the coolant temperature coefficient (CTC) in a sodium-cooled fast reactor (SFR) for a better self-stabilization of the reactor system. Previous studies on the FAST application showed that the FAST causes power excursion and temperature oscillations under anticipated transient without scram (ATWS) conditions. This study explains the variables affecting the FAST behavior and oscillations with mathematical expressions. In addition, a damping FAST is proposed to passively remove oscillations without any mechanical system. The damping FAST increases the damping ratio by reducing the pin diameter at a height where the difference between the FAST bottom elevation and the equilibrium position is maximum. The performance evaluation for the damping FAST is conducted in an innovative sodium-cooled fast reactor (iSFR) under ATWS conditions. The 7-mm damping FAST unilaterally inserts the negative reactivity in unprotected transient overpower event, dramatically lowering the core power without oscillations. While the previous FAST shows oscillatory behavior repeatedly approaching the coolant boiling temperature, the damping FAST stably keeps the coolant temperature below 800 °C. Moreover, the damping FAST lowers the maximum coolant temperatures for other ATWS events. These findings show the potential of the FAST system to significantly improve the inherent safety of SFRs.

On the use of time-dependent success criteria within risk-informed analyses. Application to LONF-ATWS sequences in PWR reactors

The classical Probabilistic Safety Analysis (PSA) does not include any time dependence explicitly. However, the success criteria (SC) could evolve during the cycle for some initiating events. In that sense, there is a type of sequence in which this time-dependency is quite important, the family of Anticipated Transient without Scram (ATWS) sequences in Pressurized Water Reactors. Therefore, a new risk-informed approach is proposed in this paper, which makes it possible to obtain the time-dependent SC evolution of the safety functions affected by the Moderator Temperature Coefficient (MTC) value. Then, the evolution of the ATWS conditional core damage probability (CCDP) could be obtained using a PSA model. To quantify the CCDP, the average values of the time-dependent failure probabilities must be computed. Finally, the comparison between the CCDP obtained through the application of the classical PSA approach and the new one makes it possible to quantify the impact of time-dependence on the SC of the headers that this new risk-informed ATWS approach can provide.

An effective method for accounting for the uncertainty of scenarios for the development of emergency situations in nuclear reactors

The main goal of the research is to develop a method for accounting for uncertain data on scenarios for the development of emergency situations in nuclear reactors of a new generation. The method involves reducing the number of deterministic analogs (ideally, to two), in which the problems of reactor optimization (mathematical programming) should be solved. The proposed method is much more efficient in comparison with traditional techniques based on the use of uniformly distributed sequences. With the same accuracy, it was possible to reduce the calculation time by an order of magnitude. As an example, the layout of the BN-800 type reactor with sodium coolant and nitride fuel is considered. The most dangerous emergency situations - ATWS (anticipated transient without scram) - are considered. The researches carried out by the author make it possible to obtain the characteristics of a new generation reactor, which are optimal for a given criterion, in which severe accidents are excluded. This is the significance of research. The method can be used to study reactors of various types. The method is relevant at the initial stages of reactor development.

Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

Review and outlook of the integral test facility PKL III corresponding studies

Abstract This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.

Assessment of dose consequences based on postulated BDBA (beyond design basic accident) A-30MWt RSG-GAS after 30-year operation

An assessment of the consequences of radiation doses due to the BDBA (beyond design basic accident) on the RSG-GAS research reactor has been done. The assessment was carried out to evaluate the KNS (Serpong Nuclear Area) EPZ (emergency preparedness zone) site after the reactor was operational for 30 years. The RSG-GAS research reactor is a 30MWt multipurpose reactor. It is the largest research reactor in Indonesia. RSG-GAS was built in the KNS Area in the Puspiptek complex which was put into operation in 1987. Previous estimations of the radiological consequences were made on accidents which were postulated based on DBA conditions. With the aging of the reactor, a study was carried out on the radiological consequences of the BDBA accident. The ATWS (anticipated transient without scram) event caused the BDBA condition which resulted in the melted of 5 fuel bundles. Source term is estimated based on an inventory of 5 melted fuel bundles, and fission products release through the reactor core, cooling system, reactor hall, and finally discharge to the environment through the reactor stack. Radionuclide inventory is calculated by ORIGEN2.1. With the influence of weather, fission products are dispersed into the air and deposited to the surface of the soil on the site. Weather and environmental data used are spatial analysis of ARC-GIS. Consequences analysis was carried out in 16 wind direction sectors within a 5 km radius using PC-COSYMA. The calculation results show the largest dose is reached in a radius below 500 m with the direction of the wind to the South. The radiation dose is below the dose limits for the exclusion and beyond exclusion area. Consequences of BDBA accident dose at RSG-GAS does not require countermeasure like sheltering, evacuation nor relocation.

Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

Transient Parameter Analysis of Non-scrammed Local Melting Accidents—A VVER 1000 Case Study

The first stage of a core degradation—based on the defense-in-depth concept of nuclear power plant (NPP) safety—is prone to fuel melting due to local blockage. The flow blockage accidents with no SCRAM happening can lead to a local fuel-clad failure, consequently affecting the safety of NPP. The present study provides an analysis of Anticipated Transient Without SCRAM (ATWS), which might lead to a condition of burning out. The accidents related to the ATWS scenarios, detailed in the case of VVER-1000/V446 reactor FSAR (Final Safety Analysis Report), include pump failure, local blockage, relative power increase, and a combination of these transients. In this research, first, drawing upon MCNPX 2.7 and COBRA-EN codes, a coupling framework is developed and then validated using an authentic reference point. The obtained results reveal that the reactor SCRAM does not occur while accidents are being investigated as there is a 10% difference in the mass flow rate reduction, a 470 kPa in the channel pressure drop, and a 204°K in the clad temperature, which constitute limitations under most pessimistic scenarios. However, under these conditions, a 70% void fraction over 12 min is observed in certain channels. Hence, burnout and local fuel melting could occur under normal operational and ATWS circumstances. According to uncertainty analyses, the occurrence of the void fraction above zero is locally definite. The transient analysis outputs could be deployed as monitoring system inputs and exploited for identifying weak points in the system.

An ultra-long-life small safe fast reactor core concept having heterogeneous driver-blanket fuel assemblies

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

Establishment of analysis methodology of RADTRAD for Maanshan PWR ATWS

Abstract In this research, the analysis methodology for Maanshan PWR nuclear power plant (NPP) ATWS (Anticipated Transient Without Scram) is established by using RADTRAD code. This methodology can evaluate the dose at Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and control room for the Maanshan ATWS. In addition, there are four leakage pathways in the Maanshan ATWS that are containment leakage, containment low volume purge, steam generator leakage, and ESF leakage. Therefore, for these leakage pathways, the RADTRAD models are established. Finally, according to the analysis results, they are below the criteria of 10 CFR 100.11 for the EAB and LPZ and Standard Review Plan (SRP) for the control room. Additionally, the steam generator leakage gives the maximum dose in these leakage pathways.

Impact of variable porosity models on the safety of high temperature pebble bed gas-cooled reactors

The Very High Temperature Reactor (VHTR) is one of the candidates of next generation of the nuclear reactors according to IAEA. Predicting the thermohydraulic behavior of High Temperature Reactors (HTR) is an important contribution to the development of this technology. Evaluation of thermohydraulic performance and a comparison with experimental data were proposed to the international research community. The impact of porosity of pebble beds in the thermohydraulic performance in normal operation and under accident condition is the goal of many kinds of research. In this sense, the three-dimensional CFD thermohydraulic simulation of a HTR using ANSYS CFX has been done. The full-scale approach is used to study the influence of different porosity scheme in the main structural materials of the HTR-10 reactor. The porous medium model was used to simulate the pebble bed core zone and the discharge system of pebbles. A comparison with experimental data was made. The steady-state thermohydraulic performance of the HTR-10 reactor at normal operation and anticipated transients without scram (ATWS) tests result, were used. In this work, was possible to estimate the effect of the porosity in the interior regions of the reactor and the impact on the temperature distribution. Finally, the study broadens the knowledge about the impact that the use of variable porosity has on safety analyses in pebble beds reactors.

Full-scale dryout/rewet instability tests in a BWR rod bundle and TRACE assessment

The regulatory purpose of the current work is to study fuel damage mechanisms and processes that were predicted to occur under certain unstable conditions that could arise during postulated anticipated transients without SCRAM (ATWS) for boiling water reactors (BWRs) operating in the maximum extended load line limit analysis plus (MELLLA+) expanded operating domain. The Nuclear Regulatory Commission (NRC) conducted experiments at the high pressure Karlstein Thermal-Hydraulic Experimental Facility (KATHY) using a full-scale simulated BWR fuel assembly. The simulated assembly includes representative features of modern BWR fuel bundles, including commercial spacers, part length fuel rods and water rods. KATHY includes two features that allowed the NRC to conduct experiments that subject the simulated fuel assembly to unstable conditions similar to those expected during postulated ATWS events. These features are the downcomer, which allows the loop to be operated under natural circulation flow, and the proprietary SINAN control program, which allows the heater rod voltage to vary in response to changing thermal-hydraulic conditions, which in turn, simulates reactivity feedback. The NRC conducted several tests which produced growing flow and power oscillations and these oscillations were allowed to proceed up to the point where the heater rod surface failed to rewet and went into the early stages of heat-up. The experimental data were analyzed to determine the temperature threshold for failure to rewet and the experimental results have been compared to various models of the minimum stable film boiling temperature (Tmin). The results show that the homogeneous nucleation temperature provides a reasonable estimate of this temperature threshold. Further, the NRC's TRAC-RELAP Advanced Computational Engine (TRACE) code was used to simulate the experiments to assess the TRACE performance for simulating failure to rewet during postulated ATWS-I events. When compared to the experimental data, TRACE produces slightly conservative estimates of the cladding temperature when an appropriately low Tmin value is used (i.e., one consistent with homogeneous nucleation). Therefore, TRACE is appropriate for predicting cladding heat-up during postulated ATWS-I events.

A Rigorous Method for Selecting the Most Limiting Exposure During Cycle for Anticipated Transient Without SCRAM with Instability Analysis

The U.S. Nuclear Regulatory Commission (NRC) staff often perform confirmatory analyses using the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes to assist in regulatory decision making. Recently, the NRC staff have performed numerous such analyses of anticipated transient without SCRAM (ATWS) with core instability (ATWS-I) scenarios for boiling water reactor license amendment requests to expand the power/flow operating domain. In the conduct of these confirmatory analyses, the staff have simulated oscillatory conditions in the reactor core under certain ATWS conditions that result in regional mode (or out-of-phase mode) power oscillations. The nature of these regional oscillations may present a challenge to fuel damage limits. Therefore, there has been interest in methods to identify the most limiting point in cycle exposure. It has been conventional wisdom that the core is most susceptible to regional mode oscillations when the fission cross section is greatest, leading to the common practice of analyzing these events at the peak hot excess (PHE) exposure point in the cycle. The staff have found some limitations in applying the PHE concept in a consistent manner. In the current work, the NRC staff have developed a more rigorous method for identifying the most limiting cycle exposure by directly considering the core flow rate, the axial power distribution, the first harmonic mode shape, and the eigenvalue separation between the fundamental and first harmonic modes. This method is a more rigorous method to screen the various exposures between beginning and end of cycle. An example case is shown to demonstrate the application of this methodology.

Unmitigated severe accident analysis for a PWR using MELCOR

After the two major nuclear accidents in history considered to be beyond-design-basis accidents (BDBA) – the Chernobyl Reactor Four explosion in Ukraine (1986) and the Fukushima Daiichi accident in Japan (2011), the United States Nuclear Regulatory Commission and other countries have included in the Final Safety Analysis Report (FSAR), a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), with significant damage to the reactor core. The Safety Analysis Report (SAR) is the most important document used by a regulatory body to assess the adequacy of plant safety at all stages of the life time of a Nuclear Power Plant (NPP), and presents the licensing basis of a plant.This study aims at comparing the progression of severe accidents during a Total Loss of Feed Water Accident (TLOFW) and a TLOFW combined with An Anticipated Transient Without SCRAM (ATWS) reactivity accident for a Pressurized Water Reactor (PWR) type NPP. This paper takes into account the results of the PSA Level 1 for the reference NPP. The calculations were performed with the deterministic computer code MELCOR version 2.2 used to analyze severe accident progression in NPP with a Light Water Reactor (LWR). The paper presents a hydrogen risk assessment – the assessment is deterministic, s a prerequisite before performing PSA Level 2. The hydrogen source is determined, too. Saphiro's diagram is applied to investigate the containment atmosphere condition (inert or combustible) for two scenarios. The main results of the study reveal that the containment integrity of the reference PWR NPP when both accidents occurred is maintained. However, the core is severely damaged. A future work will investigate mitigation strategies to limit the hydrogen risk during accident.