Severe Accident Sequences Research Articles

Modelling nuclear fuel behaviour with TAF-ID: Calculations on the VERDON-1 experiment, representative of a nuclear severe accident

The chemical behaviour of the main elements present in a PWR irradiated fuel sample (UO2 fuel, Zr from the Zircaloy-cladding, Pu, Np, and fission products such as Ba, Ce, Cs, Gd, I, La, Mo, Nd, Pd, Rh, Ru, Sr, Tc, Te, etc.) submitted to a nuclear severe accident type sequence (VERDON-1 experiment) has been investigated by thermodynamic calculations using the TAF-ID database. Particular emphasis has been placed on the chemical behaviour of the fission products Ba, Cs, Mo, and Zr, and the interactions between the Zircaloy-cladding and the UO2 matrix. Calculation results have been compared to experimental observations completed during the VERDON-1 experiment (instantaneous release of fission products and final chemical state of elements in the sample). Results presented in this manuscript do not account for non-equilibrium phenomena that have a major impact on fission product behaviour (e.g., chemical diffusion, mass transport induced by temperature gradients in the fuel pellet, etc.). None the less, these TAF-ID calculations have described accurately the melting of the sample following the formation of a mixed (U,Zr)O2-x phase because of the interaction between the fuel and the Zircaloy-cladding. Furthermore, calculations have assisted in the explanation of the instantaneous released fractions observed for Ba and Mo during the experiment (as a function of the experiment atmosphere and temperature).

Investigating the effectiveness of outfitting nuclear power plant with automatic seismic trip system based on the early detection of earthquake

Safety analysis indicates the significant effect of seismic induced accidents in overall risk of nuclear power plants located in those quake-prone regions of faults vicinity. In other words, since earthquake impacts the reliability of safety functions, seismic induced Core Damage Frequency is of considerable importance in total CDF of NPPs. One of the main concerns in seismic safety is the ability of plant safety systems to immediately shutdown the reactor and reach a safe shutdown state, upon earthquakes of predefined accelerations. Albeit, consensus on whether to implement an automatic trip signal upon earthquake of predefined acceleration, or not, is a controversial issue, with ambivalent interpretations of nuclear safety and electric power dispatching bodies. This article intends to perform a so-called risk-informed-decision-making study to decide on provisioning of Reactor Protection System of an, with automatic digitized seismic trip signal, in addition to the existing trip signals. In this regard, the two most severe accident sequences of Loss of Offsite Power and Loss of Coolant Accident are chosen to investigate and compare the advantageous and disadvantageous of early trip of the reactor on non-destructive primary wave before arrival of the potentially damaging secondary wave. Both deterministic behavior, from MELCOR model of typical PWR plant, and the probabilistic measures, from Risk Spectrum model of Level 1 Probabilistic Safety Assessment, are taken into consideration to analyze the accident progression for defined sensitivity cases on peak ground acceleration (pga) level of actuating the seismic trip setpoints.The results are useful to reach an agreement for pros and cons of outfitting NPPs with Automatic Seismic Trip System. Based on the results, an optimized actuation level of pga for seismic trip setpoint can be determined from the risk point of view. Investigating the late and early effects of the proposed change on the response of the plant will be useful in accident management decision making.

A three-dimensional neutronics – Thermalhydraulics Unprotected Loss of Flow simulation in Sodium-cooled Fast Reactor with mitigation devices

Within the framework of the Generation IV Sodium-cooled Fast Reactors R&D program at CEA, a low sodium void worth core concept has been developed. In this context, the core behavior in case of severe accidents is being assessed. Such transients are simulated with the SIMMER mechanistic code which gives reference simulation of the accidental transient. In this context, this paper presents the reactor case calculation in a heterogeneous ASTRID like core realized with a preliminary version of SIMMER-V code. Recently, significant effort on SIMMER code development and validation was realized by CEA and JAEA. The progress realized on the SIMMER code allows us to test the code behavior to simulate the whole severe accident sequence in three-dimensional geometry. After the calculation of steady state, the accidental transient of type of Unprotected Loss of Flow is being explored with SIMMER-V Preliminary Version (PV). A principal point of this study was to investigate the reactivity effects with fuel relocation under three-dimensional core representation including the mitigation devices. The calculation has indicated that the fuel discharge from the core via mitigation devices connecting the core zone with the core catcher is possible and may eliminate further reactivity increase. The main insight of the 3D studies in term of corium discharge efficiency is highlighted in the paper.

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

ABSTRACT In order to improve LWR source term under severe accident conditions, the first version of a fission product chemistry database named ‘ECUME’ was developed. The ECUME is intended to include several datasets of major chemical reactions and their effective kinetic constants for representative severe accident sequences. It is expected that the ECUME can serve as a fundamental basis from which fission product chemical models can be elaborated for use in the severe accident analysis codes. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system from 300 to 3000 K. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate fission product chemistry in Cs-I-B-Mo-O-H system under LWR severe accident conditions, where kinetics effects should be considered.

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Core Degradation and material migration research is one of the key steps in severe accident research. A review of core degradation and melting materials immigration research in light water reactor is important to encourage relevant research. Relevant experiments and numerical analyses are reviewed in this paper. There have only been a few experiments on severe accidents performed due to their high cost. They focus on different aspects of core degradation and material migration, including early stage and late stage accidents in both PWRs and BWRs. All the current experimental data should be fully utilized due to the limited data available. On the other hand, the data available from numerical analyses for severe accidents is very extensive. There are already many systemic severe accident codes developed by different organizations. They provide severe accident sequences analysis, severe accident prediction and are also important in security policy formulation. At the end of this paper, relevant work in Xi’an Jiaotong University are introduced and these works are expected to reveal parts of severe accident relevant research in China.

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty. The probability for steam generator tube creep rupture is estimated at 0.17.

MELCOR severe accident analysis for a natural circulation small modular reactor

A natural circulation small modular reactor (NCSMR) is designed and the severe accident of the NCSMR was analyzed with MELCOR code in this work. The NCSMR is a 330 MWt light water reactor no reactor coolant pumps and no active safety injection systems similar to the design of NuScale reactor. A MELCOR code model was developed for the reactor. The MELCOR model shows consistent results in steady state operation with the design values. The NCSMR severe accident analyzed in this work was induced by the small break loss of coolant accident in the reactor pressure vessel and a crack in the containment pool wall. The decay heat removal system is assumed to fail for the main steam line break. The results indicated that the RPV pressure decreases quickly while the containment pressure increases followed by the opening of reactor vent valves. The core exposes and heat up with the coolant leakage and finally collapsed to the lower plenum, leading to the lower head breach. High temperature core corium quickly broke the relatively thin containment vessel wall, and was discharged to the containment pool floor, leading to the molten corium concrete interaction. The fission products release during the accident process was also analyzed. The severe accident sequence shows that the NCSMR has much longer time duration for the reactor vessel lower head failure to occur compared to conventional commercial nuclear power plant under severe accident conditions. The analysis results help identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents in NCSMR.

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

Strategy validation of severe accident management guidance for Qinshan-I NPP under typical severe accident sequence

Strategy validation of severe accident management guidance for Qinshan-I NPP under typical severe accident sequence

A comparison of fission product behavior for typical PLWR and PHWR with FCVS operation

Since the accident at the Fukushima Dai-ichi nuclear facility, a severe accident source term analysis has been recognized as a very important research area for accident management and emergency planning. In this paper, the fission product behaviors of a severe accident sequence were analyzed for a typical pressurized light water reactor (PLWR) and a pressurized heavy water reactor (PHWR). The OPR-1000 and CANDU6 were selected as the reference plants. A station blackout scenario, in which all off-site power is lost and the diesel generators fail, is simulated as the initiating event of a severe accident sequence. The results obtained are for (1) the fission product release from the reactor core, (2) the fission product distributions in the reactor coolant system, molten corium, reactor vessel, or containment building, (3) the fission product release from the containment building, and (4) the comparative characteristics of the fission product source term depending on the operation conditions of a filtered containment vent system.

Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)

Abstract In the frame of the nuclear safety research program of the Helmholtz Association HZDR performs fundamental and applied research to assess and to reduce the risks related to the nuclear fuel cycle and the production of electricity in nuclear power plants. One of the research topics focuses on the safety aspects of current and future reactor designs. This includes the development and application of methods for analyses of transients and postulated accidents, covering the whole spectrum from normal operation till severe accident sequences including core degradation. This paper gives an overview of the severe accident research activities at the Reactor Safety Division at the Institute of Resource Ecology.

Study on Heat Removal Capacity of PCCS of Advanced Passive PWR during Severe Accidents

In the advanced passive pressurized water reactor, a passive containment cooling system (PCCS) has been adopted to cool the containment—comprising a cylindrical steel vessel—during postulated accidents, whereby the decay heat is removed through water film evaporating enhanced by air cooling outside the containment. In this study, an integrated safety analytical code is used to study the heat removal capacity of PCCS during severe accidents and its influence on severe accident management measures. The coupled analytical model includes the reactor cooling system, engineered safety features, containment system, and PCCS. Containment responses during typical design-basis accidents and integrated severe accident scenarios are calculated and validated using a design control document and probabilistic risk assessment, respectively. Four typical severe accident sequences that contribute to core damage frequency or containment high pressure are selected to evaluate the containment response. The results show that the containment pressure can be controlled at a relatively low level within 72 h with the heat removal by PCCS. Analysis of the effects of PCCS water cooling recovery during the late period of the accident sequence in severe accident management guidelines alerts as to the risk of hydrogen combustion after breaking the steam-inert atmosphere inside containment. Moreover, sensitivity analysis has been performed to study the influence of the water film coverage rate and environmental air temperature, and it shows that a decrease of the water film coverage rate and an increase of the environmental air temperature reduce the PCCS cooling capacity.

Evaluation of RCS injection strategy by normal residual heat removal system in severe accident management

Severe Accident Management Guidelines (SAMGs) suggests mitigating the consequence of severe accident scenarios by using the non-safety systems if the safety systems are unavailable. For 1000MWe advanced passive pressurized water reactor (PWR), the normal residual heat removal system (RNS) is proposed to implement the Reactor Coolant System (RCS) injection strategy during severe accidents if safety systems fail. Therefore, evaluation of the effectiveness and negative impact of RNS injection strategy is performed, in which two typical severe accident sequences are selected, which are the typical low-pressure core melt accident sequence induced by Large-break Loss of Coolant Accident (LLOCA) with double-ended guillotine break at cold leg and the typical high-pressure core melt accident induced by Loss of Feed Water (LOFW), to analyze RCS response using the integrated severe accident analysis code. The plant model, including RCS, Engineering Safety Features (ESF), containment and RNS, is built to evaluate the effectiveness of RNS injection by comparing the sequences with and without RCS injection, which shows that RNS injection can terminate core melt progression and maintain core cooling in these accident sequences. However, hydrogen generated during the core reflooding is investigated for the negative impact, which shows that RNS may increase the hydrogen concentration in the containment. For the sequence induced by LOFW, two different RCS depressurization measurements are compared, which shows that opening ADS stages 1–3 valves can cause a rapid increase of hydrogen concentration in the In-containment Refueling Water Storage Tank (IRWST) compartment and may lead to hydrogen detonation risk if hydrogen igniters are unavailable, while hydrogen can be well dispersed into the containment by intentionally opening ADS stage 4 valves, which is suggested as the preferred measure.

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.

ICONE23-2085 Parametric Study for Impact of In-Vessel Chemical Forms of Cesium and Iodine on Source Term and pH of Aqueous Phase

In order to evaluate the effect of chemical forms of Cs and I on source terms and solution pH in a severe accident of a BWR, parametric analysis was performed with an integral severe accident code, THALES2, developed by JAEA. In the present analysis, THALES2 code was modified to take into account CsOH, Cs_2MoO_4 and CsBO_2 as Cs chemical forms, and CsI and HI as I chemical forms. The severe accident sequence examined was similar to that occurred at unit 3 of the Fukushima Dai-ichi NPP. Due to the effective scrubbing, approximately 90% of the initial core inventory of Cs and I was predicted to be retained in the water pool of the suppression chamber, resulting in limited influence of Cs chemical form on Cs source term. On the other hand, the present analysis indicated that solution pH of the water pool was strongly affected by chemical forms of Cs and I. This outcome implies that chemical forms of Cs and I influence I source term since the formation of volatile I species such as I_2 and organic iodine in the water pool depends strongly on solution pH.

가압중수로형원전의 중대사고 대응능력 연구

The realistic cases causing severe core damage should be analyzed and arranged systematically for preparing an accident management of the specific nuclear power plant. The objective of this paper is to establish basic technical information for reactor safety and reactor building integrity management strategies in CANDU reactor severe accident. For the development of severe accident management strategies, plant specific features and behaviors must be studied by detailed analysis works. This analysis scope will serve to cover overall methods and analyzing results to understand the reactor building integrity status in the most likely severe accident sequences that could occur at CANDU reactor. Also analysis results could help prevent or mitigate severe accidents for the identification of any plant specific vulnerabilities to severe accidents using the probabilistic safety assessment (PSA) quantified results.

In-vessel retention coolability evaluation for Chinese improved 1000 MWe PWR

After the Fukushima Daiichi accident, the Chinese National Nuclear Safety Administration requires enhancements of safety and mitigation capability under severe accidents for nuclear power plants (NPPs). In-vessel retention (IVR) strategy is proposed to mitigate the consequences of severe accidents for pressurized water reactors (PWR) to satisfy the requirements. Therefore, a proposal with passive cavity injection system (PCIS) is designed for Chinese improved 1000MWe PWR for implementation of IVR strategy. Besides, to condense the steam generated in the cavity head, a passive heat exchange system (PHES) is also designed for passive containment cooling. To evaluate the coolability of the PCIS, a computer code is developed to assess the effectiveness of IVR strategy based on the integrated risk oriented accident analysis methodology (ROAAM) approach. In order to obtain input parameters for assessment, seven severe accident sequences are screened based on Level 1 PRA results and analyzed with integral accident analysis computer code. With the characteristics of the molten pool and key parameters gained from accident analysis, the IVR assessment program is developed with MATLAB code considering parameters of decay power, zirconium oxide fraction, the mass of stainless steel (SS). The influence of thermophysical parameters, critical heat flux (CHF) correlation and mass of metal layer on IVR effectiveness are discussed as sensitivity study for the assessment. Thermophysical parameters have the major effect. Combining the sensitivity study, assessment results show that the IVR strategy for Chinese improved 1000MWe PWR has an acceptable high success probability. However, for engineering practice, detail system design and analysis should be evaluated in the near future.

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.

Material Effect in the Nuclear Fuel–Coolant Interaction: Analyses of Prototypic Melt Fragmentation and Solidification in the KROTOS Facility

During a severe accident sequence in a pressurized light water reactor, the hot (~3000 K) molten materials (corium) coming from the degraded reactor core may generate a violent interaction if they come in contact with water. This melt-water interaction, called fuel-coolant interaction (FCI), may damage the structures and threaten the reactor integrity if there is a steam explosion. FCI occurs generally in two phases: a premixing phase, during which the molten corium jet is fragmented into large droplets and mixed with the water, and the explosion phase, during which the vapor film that has developed around the fuel droplets is destabilized and the droplets are finely fragmented. The presented work covers a set of experimental studies describing the morphology and nature of the solidified materials after interaction with water. Debris from experiments performed in the KROTOS (Commissariat à l’Énergie Atomique, Cadarache, France); PREMIX, ECO (Karlsruhe Institute of Technology, Karlsruhe, Germany); and MISTEE (Royal Institute of Technology, Stockholm, Sweden) facilities have been characterized by metallographic, analytical, and microscopic techniques. These post-test analyses are able to provide important information on the solidification path and other main phenomena involved during FCI. It was found that the behavior of metallic and oxide melts differs significantly from the standpoint of debris morphology. Oxide melts that underwent simple coarse fragmentation showed spherical or angular rocklike shape, unlike metallic melts. A statistical analysis was performed on the debris from the KROTOS tests; a data set of particles was described by the circularity, solidity, and porosity. The mechanism of water ingression (Kim and Corradini) inside the melt droplet was observed to be the key mechanism of fine (secondary) fragmentation. The particles participating in fine “thermal” fragmentation have significantly higher porosity, up to ~30% for prototypic corium in the KROTOS facility. It was calculated that only a part of the premixed melt participates in fine fragmentation, i.e., about one-third of the melt mass for the KROTOS tests using UO2-ZrO2 mixture.

Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters.The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25K/s); (2) oxidation phase (the cladding temperature about 1800K in hottest region, steam mass flow rate 2g/s); (3) bottom flood phase (characteristic cooling time about 600s, water mass flow rate 10g/s).SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.