Core Damage Progression Research Articles

Radiological impact and emergency zones for small iPWR with different approaches for source term calculation

Assessment of the potential radiological impact and emergency zones for small Integral Pressurized Water Reactors (iPWR) is very important for their employment in the near term. In this work, an attempt is made to provide an analysis and estimation of the envelope of the potential impact from a severe accident in an iPWR. The source term is calculated from the solution of the lumped aerosol concentration equation in the containment, by using different approaches for estimating the various parameters. The first approach comprises a straightforward use of the methodology and parameters used for large reactors differing only in the lower power level of an iPWR. To account for the anticipated enhanced aerosol retention in the containment and the slower core damage progression in case of iPWRs, two additional approaches are then used. The values of the deposition velocities required are specified using relevant data from available detailed computational and experimental studies. Conservative, yet representative, accidental conditions in the containment are specified, based on insights offered by available analyses with severe accident codes available in the open literature and disclosed regulatory safety assessments. The radiological impact is assessed using a Gaussian plume model, whereas the exposure is assessed against the latest reference levels and dose criteria. Fetus exposure was found to determine the level of the radiological impact during the plume phase. The radioactive releases and the associated radiological impact may differ decisively depending on the approach used to estimate the source term. Consideration of the enhanced safety attributes of iPWRs (namely, enhanced radioactivity retention in the containment and slower accident progression) leads to a reduction of more than one order of magnitude, from about 75TBq to 4TBq and 6TBq to 0.3TBq, in terms of I131 and Cs137 releases to the atmosphere, respectively. Correspondingly, emergency distances are estimated to span from about 4 km - a level commensurate with the generic distances suggested in international standards for low power nuclear reactors - to a distance as low as 600 m if the favorable characteristics of iPWRs are considered, thus implying that no significant measures are required off-site.

CSPACE for a simulation of core damage progression during severe accidents

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

Morphology and phase distributions of molten core in a reactor vessel

Investigations on the morphology and phase equilibrium characteristics of corium were performed by a series of experiments in parallel with thermodynamic phase equilibrium analyses. Melting and solidification experiments were performed using corium consists of U, Zr, ZrO2, SS, and B4C. TROI-49 and TROI-50 experiments with corium compositions representing Pressurized Water Reactor (PWR), whose compositions are similar to those of MA-3 and MA-4 of OECD MASCA (Material Scaling) Program while amount of mass is 5 times more, resulted in two-layered structure with oxide rich layer on top of metal rich layer. Experiments of FK-1 and FK-2 with a mixture of UO2, ZrO2, Zr, SS and B4C at a representative Fukushima Daiichi Nuclear Power Plant (FDNPP) corium composition resulted in a formation two-layered structure of corium with upper layer rich in metal and lower layer rich in oxide. Zirconium diboride phase trapped boron and mainly distributed in the metallic layer. Predictions by thermodynamic calculations for the equilibrium phase distribution and chemical compositions of each phase using the NUCLEA thermodynamic database with a focus on U-Zr-O-Fe system were in good agreement with the experimental results. Agreement in terms of U/(U+Zr) was quite close while there were dispersions in oxygen and steel components. The temperatures for the formation of two immiscible liquids calculated by the NUCLEA for FK-1 and FK-2 was 300 K higher than the experimental observation. This point has to be looked at for improving the NUCLEA models for corium containing B4C, by introducing a non-zero boron solubility in the ceramic phase. The agreement between the NUCLEA predictions and the results of experiments clearly indicate that thermodynamic equilibrium phases play an important role in governing the core damage progression, which is important not only for the severe accident management but also for decommissioning and defueling process for the Fukushima Daiich damaged reactors.

An analysis on the consequences of a severe accident initiated steam generator tube rupture

An analysis on the thermal hydraulic behavior, core damage progression and radiological releases of a steam generator tube rupture event initiated by a station black out event for APR1400 is provided in this paper. The analysis was performed by using MELCOR, where counter-current natural circulation, mixing in the steam generator inlet plenum, creep ruptures of reactor coolant system components and U-tube, and mitigation actions by the operator are modeled. A station black out accident with an opening of atmospheric steam dump valves (ADVs) at the time of severe accident management strategy entry condition with a failure of a supply of auxiliary feed water to the steam generators selected as a base case resulted in a creep rupture of U-tubes with a direct release of radioactive materials into the atmosphere. It was shown that a mitigation action by the operator for an aggressive cooldown of the reactor coolant system using ADVs and supplying the water to the steam generator even after the creep rupture of U-tubes was effective. It arrested core melt progression and significantly reduced the amount of radiological releases to the atmosphere.

Safety insights from forensics evaluations at Daiichi

Although it is clear that the accident signatures from each affected unit at the Fukushima Daiichi Nuclear Power Station [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents.Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This paper reports initial results from the US Forensics Effort to utilize examination information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. In this paper, three examples are presented in which examination information, such as visual images, dose surveys, sample evaluations, and muon tomography examinations, along with data from plant instrumentation, are used to obtain significant safety insights in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights confirm actions, such as the importance of water addition and containment venting, that are emphasized in updated guidance for severe accident prevention, mitigation, and emergency planning.

Modeling and verification of three-dimensional simulation for BWR in-vessel core degradation

A Multifunction Model has been developed for describing BWR in-vessel core degradation from initial cladding temperature rise to eventual molten materials accumulation in the lower part of reactor vessel as a fundamental safety research by Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R). The Multifunction Model aims to reduce the uncertainties of severe accident assessment related to physical models on core damage progression, melt formation and material relocation processes in the reactor vessel. The Multifunction Model consists of a thermal-hydraulic module, a fuel pin behavior module and a neutronic module with three-dimensional geometry. For reducing the uncertainties, the thermal-hydraulic module incorporated multi-phase, multi-component and multi-velocity field with multiple chemical and eutectic reactions. Four fuel failure models were incorporated into the fuel pin behavior module in order to simulate the failures of cladding rupture, brittle fracture, cladding melting and pellet dissolution. In order to decrease the calculation time, the data table of macroscopic cross-section was prepared for neutronic diffusion calculation of three fuel conditions in advance. The verification results revealed that the Multifunction Model demonstrated a reasonable capability to simulate the BWR in-vessel core degradations. The validation will be carried out for existing severe accident experiments.

B244 安全向上策に関わる技術課題3 : 圧力容器の健全性(OS3 軽水炉・新型炉・原子力安全(6))

Core damage and melt progression behavior due to loss of core cooling function is one of the important factors in the severe accident. After TMI-2 accident, some integral severe accident analysis codes have been developed; however, the core damage behaviors, the relocation behavior of molten core debris and failure modes of reactor pressure vessel have still large uncertainty. The Atomic Energy Society of Japan (AESJ) developed a new thermal-hydraulics safety evaluation fundamental technology enhancement strategy roadmap (TH-RM) for LWR safety improvement and development after Fukushima Daiichi NPP accident. This paper describes the major technical issues and activities to secure the integrity of reactor pressure vessel which are discussed in the TH-RM developing process.

Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding.

Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core. This article presents the analysis of large loss of coolant accident (LOCA) with emergency core cooling system failure and station blackout cases without any operator intervention and when the cooling of reactor control rods is restored. For the station blackout case the other non-regular measures (depressurisation of reactor cooling system, water injection from hydro-accumulators, deaerators and city water system) are evaluated. The analysis was performed using RELAP5 Mod3.2, RELAP/SCDAPSIM and RELAP5-3D codes. The use of three codes allows to investigate all aspects of accident with long-term loss of heat removal from the core: RELAP5 code allows to perform thermal-hydraulic analysis using detailed RBMK-1500 model; RELAP/SCDAPSIM code is necessary for modelling of core damage progression; RELAP5-3D code with special “multidimensional heat conduction” model allows to model heat transfer from hot fuel channels through graphite column in radial direction. Results of the analysis have shown, that the heat removal from control rods cooling circuit allows to slowdown effectively the core heat-up process.

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

The impact of assessing lower head integrity on boiling water reactor probabilistic risk assessments

Abstract Over the last several decades, much effort has been directed at estimating the likelihood of a large early release of radioactivity during a nuclear accident. This effort has culminated in the Individual Plant Examinations (IPEs) for the over 100 US nuclear power plants and the NUREG 1150 study. The large early release of radioactivity requires core damage with loss of primary containment integrity during the accident. Given a successful reactor scram, early containment failure coupled with a large release of radioactivity will only occur if the reactor core vessel is breached by core debris. Most IPE/PRA studies performed to date have not considered the possibility of quenching core debris in the lower plenum. Consequently, lower head failure is presumed to closely follow the onset of core damage. Therefore, these assessments did not address the role that in-vessel debris retention plays in preserving primary containment integrity, nor do they propose a criterion for evaluating the integrity of the vessel lower head given that core damage has occurred. Yet preserving the vessel lower head integrity is a necessary condition for satisfying the plant design and licensing basis. Therefore, a more complete treatment of the risk associated with nuclear plant operation includes an evaluation of the ability to retain the core debris in-vessel. This paper presents a performance requirement for vessel integrity to be used in probabilistic risk assessments; evaluates the impact the core damage progression and lower plenum quenching models have on the likelihood of terminating the damage progression in-vessel; documents the significant reduction in BWR containment failure probability that can occur when appropriate core damage and lower head quenching models are used; reviews the implications of core debris quenching in the lower head on BWR PRA modeling; argues why crediting the capability to maintain vessel integrity is necessary from a safety point of view. These results and conclusions are derived from consideration of a BWR 4 plant with a 251 inch vessel. However, the concepts are generally applicable and results specific to other BWR designs can be developed using the methodology presented in this paper.

DF-4 Analysis Using SCDAP/RELAP5

In this paper the SCDAP/RELAPS severe accident analysis computer code, developed at the Idaho National Engineering Laboratory, is used to analyze the fourth in a series of debris formation experiments. The debris formation-four (DF-4) experiment deals with heatup and meltdown of a boiling water reactor (BWR)-representative fuel and control blade assembly segment, performed in the Annular Core Research Reactor at Sandia National Laboratories. The DF-4 experiment provides data that are used to validate core damage progression and BWR-specific models to gain an understanding of the phenomena occurring in the bundle during a severe BWR accident and to identify additional modeling needed in severe accident codes. The SCDAP/RELAPS model used for this analysis accurately predicts the key damage events, which include control blade melting, channel box relocation and runaway oxidation, the order and timing of these events, and the maximum bundle temperature. From these analytical calculations, an accident scenario and insights into phenomena occurring during a severe BWR accident are developed.

Materials Interactions and Temperatures in the Three Mile Island Unit 2 Core

AbstractCore debris samples obtained from different regions of the Three Mile Island Unit 2 (TMI-2) core were examined to characterize the interaction among core components and the coolant, to determine the peak temperatures at which the interactions occurred, and to evaluate core melt progression in TMI-2. Estimates of peak temperatures were needed from these samples because of the strong influence that temperature has on core damage progression and fission product behavior. The peak temperatures can be bounded by comparing the observed microstructure and compositions with established phase diagrams. The microstructures were determined by optical metallography and scanning electron microscopy, and compositions were determined by energy and wavelength dispersive X-ray spectroscopy and scanning Auger spectroscopy.The material interactions among the core components are very complex and involve not only the interaction between the Zircaloy cladding and the UO2 fuel, but interactions with control rod material...

TMI-2 accident scenario update

Estimates of the end-state core configuration have been developed from recent inspection of the lower regions of the TMI-2 core, core support assembly, and lower plenum regions. The inspection data have provided a basis for estimating the extent of damage to the core and core support structures and have confirmed that the migration pathway of the molten core material to the lower plenum occurred in the east quadrant of the reactor vessel. This paper integrates the core inspection data with other TMI-2 data and supporting analysis to update the best-estimate core damage progression scenario.

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min. The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination. The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident. Results indicate that: 1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench. 2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO 2 pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid UZrO alloys could have formed. 3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench. 4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating. Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.