Design Basis Accident Research Articles

A best estimate plus uncertainty safety analysis framework based on RELAP5-3D and RAVEN platform for research reactor transient analyses

This paper develops a best estimate plus uncertainty (BEPU) framework for research reactor transient safety analysis. The BEPU framework is developed based on the system level reactor safety analysis code RELAP5-3D and the data analysis platform RAVEN developed by Idaho National Laboratory. Within the framework, a sensitivity analysis procedure is first conducted to identify the contributions and ranks of individual input parameters to the user-defined figures-of-merit (FOMs) associated with specific transient phenomena. An uncertainty analysis procedure is then performed to quantify the uncertainties of the FOMs resulting from the uncertainties of the input parameters. Many useful outcomes can be realized through the BEPU analysis. Specifically, the sensitivity information obtained from the sensitivity analysis will provide insights about the influence of each different input parameter on FOMs. The uncertainty information obtained from the uncertainty analysis will imply the range of response deviations caused by the propagation of errors existing in various input components.As a case study for research reactors, the developed BEPU framework was employed to perform design-basis accident (DBA) analysis for one conceptual research reactor design proposed at the National Institute of Standards and Technology (NIST). Two hypothetical DBA scenarios, namely the reactivity insertion accident (RIA) and the loss of flow accident (LOFA), were modeled and analyzed through the BEPU framework. To demonstrate the value of the BEPU framework, the BEPU analysis results were compared to that obtained from the conventional transient safety analysis procedure, which was conducted by using commonly used transient safety analysis codes including RELAP5-3D and PARET. The comparison shows that the BEPU analysis is capable of providing additional sensitivity and uncertainty information that help confirm safety margins of the NIST conceptual research reactor during both RIA and LOFA situations, which justifies the advantages and benefits of the BEPU safety analysis framework developed in this work.

Safety cases for design-basis accidents in LWRs featuring passive systems Part 2 – Numerical investigations

This paper deals with the improvement and validation of numerical tools for the simulation of design basis accidents in nuclear power plants equipped with passive safety systems. Numerical models are implemented in the framework of the 1-D thermal–hydraulic system code ATHLET developed by GRS. Experimental reference data for the validation were obtained at the INKA test facility, a model of the KERENA reactor, reproducing the passive safety systems nearly at full scale.The validation effort focuses firstly on the accuracy of the models for the single passive components, and secondly on the ability of the numerical simulation to reproduce the interaction of all components of the KERENA design under realistic conditions as reproduced in the INKA test facility. Thermal-hydraulic models are presented and validated for two passive components of the KERENA reactor: the passive pressure pulse transmitter and the pressure-controlled flooding valve. Finally, the full model of the INKA facility, including these and other passive components, is discussed and numerical results for simulations reproducing three different design basis accidents are validated by comparison with the corresponding experimental data.

Results of Structural Analysis of WWER-1000/320 Containment Behavior under Severe Accidents

The paper presents the review and evaluation of the containment prestressing system within reinforced concrete structures under seismic loads and severe accidents. Given the complex design of the containment, the detailed finite element model has been developed and used to describe real containment behavior. Containment stress and strain state was calculated by modern LIRA software. The first stage analyzed the results of WWER-1000/320 containment stress and strain state calculation under a combination of loads caused by maximum design basis accident (MDBA) and safe shutdown earthquake (SSE) and defined minimum acceptable tension of tendons. The research determines the minimum acceptable tension of tendons in the containment prestressing system, and evaluates the strength and reliability of containment structures under a combination of loads in normal operation + design-basis accident + maximum design earthquake (NO + DBA + MDE). The verification calculations have been performed using tendon tension of 780 ton-force in the cylindrical part of the containment and 760 ton-force in the containment dome.
 The second stage covered the analysis of severe accident parameters (pressure and temperature) and the results of calculation. Stress and strain state in ZNPP-1 containment has been calculated, parameters (pressure and temperature) under which the containment can loss its protective and isolation functions have been identified, calculation results have been analysed and conclusions of containment structural integrity and ensuring the implementation of the design confining functions have been made. Based on the calculation results, it can be concluded that strength of the containment cylindrical part during a beyond design-basis accident cannot be ensured under parameters t (temperature) = 120°С, p (pressure) = 0.6 MPa.

Development of phenomena identification and ranking table for Westinghouse lead fast reactor's safety

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-size, passively safe, economic, Gen-IV nuclear reactor. An important effort within the Westinghouse LFR program is the development of the safety analysis methodology, which comprises computer code development, model development, and experimental testing. A key initial task associated with the development of the safety analysis methodology is the identification of processes and phenomena that affect the plant's capability to meet selected safety performance indicators. This is accomplished through the development of a Phenomena Identification and Ranking Table (PIRT) for selected accident scenarios which, for this specific PIRT effort, included selected postulated design basis accidents and hypothetical beyond design basis accidents in LFRs.This paper describes the role of PIRT in the development of the Westinghouse LFR safety analysis methodology and the process used in the PIRT development. Specifically, the Westinghouse LFR PIRT assessed importance of pertinent phenomena and identified gaps in their knowledge-base by evaluating current modeling capabilities and data available for validation. The ultimate goal was to provide guidance on computer code development and validation efforts and to prioritize testing to support LFR design and licensing. The key phenomena and processes that are deemed highly important for the safety performance indicators, but for which the state of knowledge is low, are presented. The testing program and analyses development are planned to address significant phenomena in the PIRT.

Assessment of designing multi-fuel filling stations

The lack of design standards for fire-prevention distances between the ground tanks of reduced hydrocarbon gases (RHG) as part of multi-fuel filling stations to facilities not related to it and on the territory (MFS) between the buildings, structures and technological equipment is the urgent problem in the design and operation of these facilities. The article evaluates the parameters of hazardous factors of possible technogenic emergencies on the technological equipment of RHG sites for workers, fuel consumers and people who are in the residential area near multi-fuel filling stations (MFS) according to the existing method. The most dangerous design basis accidents, in which the largest amount of flammable gas and gasoline is released into the environment at the MFS territory, are considered. Technical solutions have been developed to eliminate the formation of a flammable environment at the facility, as well as measures to prevent the formation of ignition sources in a combustible environment.

Fuel for VVER and PWR: Current Status and Prospects

The Rosatom State Corporation activity in the nuclear fuel market requires the development of new designs and modifications meeting the current requirements of safety and operational reliability. The history of the development of a competitive fuel meeting heightened requirements of economic efficiency, safety, and operational reliability for NPP with VVER and PWR is presented. Ways to improve the Russian fuel for homegrown and foreign reactors are described. The development of fuel resistant to beyond-design basis accidents is considered.

Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

Research and development of a transient thermal–hydraulic code for system safety analysis of sodium cooled fast reactor

A transient thermal–hydraulic code THCSA is developed to analyze the primary coolant system for pool-type sodium cooled fast reactor which includes reactor core, intermediate heat exchangers, pumps etc. The mathematical models of various components of the primary sodium circuit, numerical method, and assistant models for sodium flow and heat transfer are provided in detail. Several typical design basis accidents of CEFR such as protected LOFA, LOHS, and RIA etc, are analyzed. The key parameters such as reactor power, mass flow rate and sodium temperature etc agree well with those of the final safety analysis report calculated by DINROS. The results show that it has a benign response to transient conditions, which proves the effectiveness of the models and numerical method developed in the present work. THCSA could be used in the system safety analysis of pool-type sodium cooled fast reactor and adopted in the development of CEFR digital simulator.

Infrared observation of ballooning and burst of nuclear fuel cladding tubes

Nuclear fuel claddings can balloon and rupture at high temperatures under internal gas pressure in case of design basis accidents like loss-of-coolant-accident (LOCA). The thermal phenomena surrounding the ballooning and cracking was investigated in a series of experiments performed using zirconium alloy cladding tubes at the Centre for Energy Research in Hungary.The experiments were recorded using a high-speed infrared camera in the intermediate infrared range and the temperatures were measured in-situ. It was found that before the rupture of the cladding tube, the ballooned cladding surface forms a bulge and heats up locally, a hot spot appears. During crack propagation, the crack tip temperature is significantly warmer than the rest of the tube. The hot spot temperature and the rate of temperature increase was determined. The infrared emission coefficient of the slightly oxidized zirconium cladding samples in the intermediate infrared range was also determined. The emissivity coefficient that fit our measurements was found to be around ε = 0.89.

Influence and analysis of a commercial ZigBee module induced by gamma rays

Many studies are undertaken into nuclear power plants (NPPs) in preparation for accidents exceeding design standards. In this paper, we analyze the applicability of various wireless communication technologies as accident countermeasures in different NPP environments. In particular, a commercial wireless communication module (WCM) is investigated by measuring leakage current and packet error rate (PER), which vary depending on the intensity of incident radiation on the module, by testing at a Co-60 gamma-ray irradiation facility. The experimental results show that the WCMs continued to operate after total doses of 940 and 1097 Gy, with PERs of 3.6% and 0.8%, when exposed to irradiation dose rates of 185 and 486 Gy/h, respectively. In short, the lower irradiation dose rate decreased the performance of WCMs more than the higher dose rate. In experiments comparing the two communication protocols of request/response and one-way, the WCMs survived up to 997 and 1177 Gy, with PERs of 2% and 0%, respectively. Since the request/response protocol uses both the transmitter and the receiver, while the one-way protocol uses only the transmitter, then the electronic system on the side of the receiver is more vulnerable to radiation effects. From our experiments, the tested module is expected to be used for design-based accidents (DBAs) of “Category A″ type, and has confirmed the possibility of using wireless communication systems in NPPs.

Analysis of beyond design basis accident with hydrogen explosion in the at-reactor spent fuel storage pool of NPP with RBMK-1000

The article evaluates the consequences of a hydrogen explosion in an at-reactor spent fuel storage pool due to the development of a beyond design basis accident with a complete long-term blackout. Two scenarios of hydrogen generation in a spent fuel storage pool with a subsequent hydrogen explosion are considered. In the first scenario, the formation of radiolytic hydrogen occurs in the undried storage pool. In the second scenario, hydrogen is formed as a result of the steam-zirconium reaction in the dried storage pool when the cooling water is supplied to the walls of the emergency storage pool. The maximum explosion pressure is calculated using the shock Hugoniot adiabat. The calculations of initial static load taking into consideration hydrogen explosion exposure for each scenario were performed using the equivalent static load method. This method consists in the switch from a dynamic task to a static one, in which the maximum deflections are equivalent. The calculation of the stress-strain state of spatial structures is carried out in the ZENIT-95 program. It is shown in the work that not in all cases of hydrogen explosion, destruction of structures occurs.

Estimation of pressure during steam explosion in the sub-reactor room of NPP with RBMK-1000

The article analyzes the scenario of a severe beyond the design basis accident with core melting at a nuclear power plant with RBMK-1000. Molten reactor core falls into the sub-reactor space filled with a steam-water mixture. The interaction of melt core and the coolant can lead to the creation of conditions for the occurrence of a steam explosion. Parameters of this steam explosion have been evaluated by Hugoniot adiabat method. Such approach considers the steam explosion as detonation phenomenon, where shock wave impacts initial multiphase mixture and initiates melt droplets fragmentation and high heat flux from melt to water. Mass, momentum and energy conservation equations for initial mixture and for equilibrium explosion products mixture lead to Hugoniot equation for the adiabat. The tangent from initial point (initial mixture) on the plane “mixture volume-mixture pressure” to the Hugoniot adiabat determines Chapman – Jouguet point, which characterizes steam explosion parameters. Our calculations have demonstrated that obtained values of steam explosions do not exceeds dangerous limits.

Tolerant fuel for VVER reactors

The main factor of destruction of fuel rods in accidents with loss of coolant is associated with the vapor-zirconium reaction occurring between the fuel rod shell and the coolant (water). Improving the reliability of fuel cells can be obtained by modifying or replacing the fuel shell, materials that do not interact with the coolant during normal operation and in emergencies. The loss of coolant accident is a design-basis accident in light water reactors. The postulated accident requires an analysis of the double guillotine break in a main primary coolant pipe, which allows the coolant to freely discharge out of the primary system into the containment of the building. In the present work, the heat stored in the active zone of the reactor, the heat capacity of the fuel and the shell temperatures are determined for ATF materials U-10Mo and U3Si2. The calculated values are compared with the usually used fuel UO2. The results showed beneficial effect of using ATF materials. The above quantities are decreased dramatically for ATF, promising to replace UO2 with ATF fuels. This reduces the probabilities of accidents and oxidation in the nuclear reactors. The calculations are done for nuclear power plant, type VVER-1200.

Limitations of traditional tools for beyond design basis external hazard PRA

Probabilistic risk assessment (PRA) is being used increasingly by the nuclear industry for safety during normal operations as well as for the protection against external hazards. Computation of total risk in an external hazard PRA is dependent on hazard assessment, fragility assessment, and systems analysis. A systems analysis for propagation of component fragilities is conducted using event and fault trees. The event and fault trees for an actual power plant can be fairly large in size, which imposes computational challenges. Hence, certain assumptions are employed for computational efficiency. These assumptions typically represent the conditions imposed during the design basis (DB) scenario. The traditional PRA tools based on these assumptions are also widely applied to perform risk assessment in the context of beyond design basis (BDB) scenarios. However, some of these assumptions may not be valid for certain BDB scenarios. In addition, the probability of dependent failures also increases in BDB scenarios due to common cause failures (CCF) which usually results from design modifications, human errors, etc. In this manuscript, a simple and a relatively more complex illustrative examples are used to show the limitation of these assumptions in the numerical quantification of risk for the case of BDB conditions. Case studies with CCF events across multiple fault trees are also presented to illustrate the effect of these assumptions when traditional approach is used in BDB risk assessment. It is shown that the assumptions are valid for the case of DB conditions but may lead to excessively conservative risk estimates in the case of BDB conditions. A Bayesian network based top-down algorithm is proposed as an alternative tool for accurate numerical quantification of total risk in systems analysis.

A quantitative risk assessment methodology based on CCDF risk curve for accidental releases from fusion reactors

Abstract Fusion reactors are expected to be more safe and environmentally friendly in the development of nuclear energy systems. It is vital to fully understand the overall radiation risk level of fusion reactors in comparison with existing nuclear energy systems. The objective of this paper is to propose a risk assessment methodology for fusion reactor accidents, which is capable of quantifying the risk level comparable to other advanced nuclear energy systems. In this paper, fusion specific accident characteristics are comprehensively analyzed in comparison with ITER and fission reactors. The postulated initiating events as well as possible environmental release routes are also presented for DEMO, and then deterministic risk assessment is applied to design basis accidents, while the coupling of deterministic risk assessment and probabilistic risk assessment is suggested for design extension conditions. The societal risk of DEMO is proposed to be presented by a Complementary Cumulative Distribution Function (CCDF) risk curve.

Defining the performance envelope of reactivity-initiated accidents in a high-temperature gas-cooled reactor

The purpose of this study is to define an envelope of possible transient conditions during two reactivity-initiated accidents (RIAs) in a modular high-temperature gas-cooled reactor (mHTGR). The accident scenarios studied were a group control rod withdrawal (CRW), a design basis accident, and a control rod ejection (CRE), a beyond design basis accident. Variance-based sensitivity analysis methods were employed to obtain realistic bounds on the reactor power and maximum temperature, energy deposition, and heating rate in the fuel during these accidents. Key reactor kinetic parameters and the heat transfer properties of the fuel and moderator were varied in a RELAP computer model of a 350-megawatt mHTGR to obtain these bounds. Sobol sensitivity indices were calculated to highlight which inputs account for the most variance in the outputs. Results showed that the kinetic properties and reactivity temperature coefficients were the most dominant inputs, while thermal properties had little impact on the outputs. Compared to previous experimental work, the fuel particle energy deposition from the simulated RIAs would not be expected to cause fuel failure in any design basis or beyond design basis reactivity accident we investigated. The predicted energy deposition in fuel kernels ranged from 98 to 331 J/g-fuel, significantly less than the expected failure threshold of approximately 1500 J/g-fuel. However, the power pulse widths and heating rates used in previous fuel safety experiments were different from the values predicted in this study for both accident scenarios. Regardless of these differences, this analysis shows that the safety margins in graphite-moderated mHTGRs for reactivity accidents are very large, and additional fuel safety tests are not expected to be necessary for this application.

TRANSIENT ANALYSIS OF SIMULTANEOUS LOFA AND RIA IN RSG-GAS REACTOR AFTER 32 YEARS OPERATION

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip. From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.Keywords: RSG-GAS, Simultaneous, LOFA, RIA, PARET

Integral molten salt reactor temperature sensitivities using Serpent target motion sampling

The Molten Salt Reactor (MSR) concept has attracted several start-up companies due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. Terrestrial Energy, Inc (TEI), for example, has designed and is seeking regulator licensing for an Integral MSR (IMSR) design that contains all primary system components within a single reactor vessel containment. This study uses sensitivity analysis capabilities unique to Monte-Carlo particle transport code Serpent to determine and evaluate fuel salt Doppler and density temperature coefficients of reactivity for an IMSR over a range of normal or casualty temperature ranges. This study contrasts the temperature coefficients of three candidate molten salt fuels (FLiBe, FLiNaK, and NaF-RbF). The unique advantages and access to insightful neutron physics using Serpent code sensitivity methods are highlighted. An evaluation of the total reactor temperature coefficient and in particular the fuel salt temperature coefficients will show that an IMSR design can be self-limiting from a design basis accident perspective.

Reactor performance and safety characteristics of two-phase composite moderator concepts for modular high temperature gas cooled reactors

Graphite moderators have an extensive historical performance record, but also feature inherent challenges for modular High Temperature Gas-Cooled Reactors (mHTGRs). Challenges with graphite include non-uniform expansion and contraction under irradiation and build-up of potential energy during the bombardment of high energy neutrons that results in a large energy release under annealing. These challenges have led to the investigation and development of alternative moderators to be utilized in mHTGRs, including beryllium- and hydride-based concepts with compositions selected for favorable moderating power and the potential for improved in-service lifetime as compared to graphite. The proposed moderators are fabricated as two-phase composites with magnesium oxide, MgO, as the radiation-stable host matrix and beryllium metal, Be, beryllium oxide, BeO, or zirconium hydride, ZrHx=1 (to account for hydrogen loss from the hydride phase during processing), as the entrained moderating phase. Here, we evaluate the reactor performance and safety characteristics of these moderator concepts relative to a graphite reference using a Ft. Saint Vrain-style fuel block. We assessed the cycle length, discharge burnup, natural resource utilization, neutron flux spectra, moderating power, moderating ratio, critical size, moderator and fuel temperature feedback, fuel cycle cost, spent nuclear fuel and high level waste radioactivity per unit energy generated, and environmental impact per unit energy generated. The results demonstrate that the advanced moderators have the potential for comparable or enhanced cycle performance to that of the graphite reference case with significantly improved performance for an optimized moderator-to-fuel ratio design. These advanced moderators are also assessed from a reactor safety standpoint for Design Basis Accidents (DBAs) including Pressurized Loss of Forced Cooling and Depressurized Loss of Forced Cooling accidents for a 350 megawatt thermal prismatic-type mHTGR. The full core thermohydraulic analysis of DBAs show that the high volumetric heat capacity of the beryllium-based moderator grants them a greater margin to fuel failure in these analyses than a conventional graphite moderated system, but the lower thermal conductivity of the beryllium-based moderators leads to longer times at elevated temperatures.

Comparison of Two Different Sized Small-Break LOCAs on the Passive Safety Injection Line Using SMART-ITL Data

Even for small modular reactors (SMRs) with all large pipes removed, a small-break loss-of-coolant accident (SBLOCA) is an important design-basis accident (DBA). Experimental simulation of the SBLOCA scenario is essential before a prototype reactor is realized. The system-integrated modular advanced reactor (SMART) is one of the SMRs developed by the Korea Atomic Energy Research Institute. An integral test loop, SMART-ITL, was also constructed to carry out several types of integral thermal-hydraulic effects tests for the prototype reactor. The SMART-ITL was designed with a preserved height, 1/7th diameter, and 1/49th area, and volume-scaling ratios. Two types of passive safety systems were equipped in the SMART-ITL: a passive safety injection system (PSIS) and a passive residual heat removal system (PRHRS). The PSIS was designed to refill the coolant in the reactor coolant system (RCS) for 72 h after an accident. Under accident conditions the PRHRS prevents overheating and overpressurization of the RCS using two-phase natural circulation. The SBLOCA on the passive safety injection line is a significant DBA that should be validated for differences in break size. In this paper, the effects of two different break sizes, 2 and 7/32 in., were analyzed in order to study the effect of the maximum and minimum mass and energy loss of the RCS. In order to simulate a clear difference between maximum and minimum mass and energy loss of the RCS, heat removal by the PRHRS was performed in the maximum break size (2-in.) accident, and heat removal by the PRHRS was not conducted in the minimum break size (7/32-in.) accident. The difference in mass and energy loss of the RCS will have a significant impact on the operation of the automatic depressurization system. Using the two extreme accident simulations, it was possible to confirm the difference in accident progression caused by the difference in break size and the characteristics of the PSIS.