TMI-2 Accident Research Articles

BWR lower head penetration failure test focusing on eutectic melting

Decommissioning work occasioned by the Fukushima Daiichi Nuclear Power Station (1F) accident of March 2011 is in progress. Severe accident (SA) analysis, testing, and internal investigation are being used to grasp the 1F internal state. A PWR system that refers to the TMI-2 accident is typical for SA codes and testing, on the other hand, a BWR system like 1F is uncommon, understanding the 1F internal state is challenging. The present study conducted the ELSA-1 test, a test that focused on damage from eutectic melting of the liquid metal pool and control rod drive (CRD), to elucidate the lower head (LH) failure mechanism in the 1F accident. The results demonstrated that depending on the condition of the melt pool formed in the lower plenum, a factor of LH boundary failure was due to eutectic melting. In addition, the state related to the CRD structure of 1F unit 2 were estimated.

Analysis of primary side bleed and feed actions for severe accident management following total loss of feed water in a Swedish PWR

Severe accident management guidelines (SAMGs) have been developed and applied in nuclear power plants (NPPs) to provide the fourth level of defense worldwide since the TMI-2 accident which occurred in 1979. The primary objective of the SAMGs is to protect remaining safety boundaries and to limit release of fission products into the environment by specific severe accident management (SAM) actions with available safety equipment. This paper is concerned with the assessment of the primary side bleed and feed (PBF) actions in the SAMGs of a Swedish pressurized water reactor (PWR). The PBF actions are realized by operations of emergency injection pumps and power-operated relief valves (PORVs), following a severe accident induced by an event of total loss of feed water (TLOFW) and temporary station blackout. In the present study, MELCOR simulations are performed to investigate the effects of the SAM actions on severe accident progression and consequences. Different sets of available equipment and actuation time due to recovery of AC power are simulated. The interest of the MELCOR analysis is focused on how these uncertain parameters of SAM actions affect thermal–hydraulic response, hydrogen generation, core relocation, reactor pressure vessel failure and release of fission products. The results demonstrate the effectiveness of the PBF strategy to prevent RPV failure in the TLOFW accident and the effects of injection timing on accident consequences. The obtained insights are instrumental for validation and improvement of the strategies and actions in the plant specific SAMGs.

Long term cooling safe shutdown performance analysis for SMART with passive safety system using MARS-KS

Passive safety systems have been adopted in reactor design to enhance safety especially in SMRs after the TMI-2 accident in 1979. The passive system has high operational reliability with natural driving forces such as gravity, natural circulation and stored energy. In recent years, development of innovative SMRs has emerged as a global R&D initiative for sustainable energy to challenge the carbon neutrality of climate change. The passive safety system makes reactor design simple and thus enhances the reliability of the reactor. As advanced reactors adopt passive safety systems, regulatory requirements and guides are required. Thus, development of regulatory guide and detailed regulatory review methodology for the passive safety systems is needed. In this paper, long term cooling safe shutdown acceptance criteria and performance analysis methodology are proposed for the regulatory standard design approval for the SMART100 integral reactor with fully passive system which has been jointly developed by KAERI in Korea and K.A.CARE in Saudi Arabia. The long term cooling safe shutdown performance analyses are performed until 72 hours to evaluate the methodology, and sensitivity analyses are also performed to provide a technical basis for the regulatory independent review for the long term cooling performance of the SMART100 passive safety systems.

Numerical simulation of zircaloy-water reaction based on the moving particle semi-implicit method and combined analysis with the MIDAC code for the nuclear-reactor core melting process

During the severe accident in PWR (Pressurized Water Reactor), local phenomena (e.g., eutectic reaction, candling effect, zircaloy-water reaction) occurring in the core have a significant impact on the core melting process. However, the traditional severe accident analysis codes cannot simulate the local phenomena accurately. To solve this problem, the Computational Fluid Dynamics (CFD) method is considered. By coupling CFD method with traditional severe accident codes, the accident sequence in the core melting process can be simulated accurately. In this study, considering the complexity of physical and chemical phenomena involved in the core melting process, the zircaloy-water reaction is selected for analysis. The zircaloy-water reaction kinetics models in the Module In-vessel Degradation Analysis Code (MIDAC) are summarized. The correlations for zircaloy temperature below 1800K are compared and the results show that the Cathcart-Pawel correlation induces a considerable overestimation of the reaction rates above 1460 K. Then based on the Moving Particle Semi-implicit (MPS) method, a numerical analysis model is developed to simulate zircaloy-water reaction. A zircaloy-water reaction kinetics correlation with higher accuracy is obtained by simulations and added to MIDAC. The TMI-2 accident is selected to evaluate the newly developed correlation by comparing with other correlations in MIDAC. The results indicate that different correlations can greatly influence the structure of corium pool in the core and the melt behaviors during reflooding and long-term cooling stage. However, it can hardly influence the relocation of debris bed to the lower plenum and behaviors of debris bed in the lower plenum.

The evolution of the U.S. nuclear regulatory process

This paper provides historical perspectives and insights on the early development of the U.S. nuclear regulatory process and its subsequent evolution towards risk-informed processes. After the landmark Reactor Safety Study (WASH-1400) and the TMI-2 accident, the U.S. Nuclear Regulatory Commission (NRC) began to use probabilistic risk assessment (PRA) methods and insights in regulatory applications as deemed necessary or useful. In 1995, the NRC adopted a policy that promotes increasing the use of probabilistic risk analysis in all regulatory matters to the extent supported by the state of the art to complement the deterministic approach. The NRC then started moving toward a much expanded use of PRAs in what is termed risk-informed regulatory approach. This paper discusses the challenges and the success stories of the use of probabilistic assessment of the risk to support and inform regulatory decisions.

Supercritical transient analysis in coupled fuel debris systems of symmetric and asymmetric geometry using integral kinetic model

Analysis of a possible criticality accident during decommissioning of the Fukushima Daiichi NPS (1FNPS) by evaluating the potential energy and dose release is important for safety. Yet choosing an appropriate method for such analysis can be challenging because of the unknown state of fuel debris inside the reactors. Past experience with the TMI-2 accident, however, suggests that fuel debris at 1FNPS could be highly heterogeneous in terms of both composition and geometry consisting of physically distinct but neutronically coupled debris regions. Conventionally, a one-point kinetic model (PKM) supplemented with deterministic method was used for transient analysis. However, the PKM fundamentally does not provide space-dependent quantities. Thus, an integral kinetic model (IKM) aided by Monte Carlo transport method is more suitable because it takes space/region-dependency into account. As a preliminary analysis for the long term effort of analyzing criticality accidents involving fuel debris, in this study an effect of surrounding debris during a supercritical transient in simple two-region coupled debris systems of symmetric and asymmetric geometry were investigated using the IKM with simple Doppler feedback mechanism. The obtained results included region-wise fission rate profile and energy release. The results indicated that total energy release decreased about exponentially as the distance between two regions increased.

Characterisation of aerosols produced in a simulated severe nuclear accident using electron microscopy

The chemical composition of aerosols formed by revaporisation and subsequent condensation of fission products and simulant material was investigated using electron microscopy. A stage dedicated only to the collection of aerosol particles on carbon coated copper grids was designed and tested. As raw material for these experiments core material from the TMI-2 accident, fission product deposits from the Phébus FPT3 tests and Cs2MoO4 powder were used. The examination of aerosols was made by transmission and scanning electron microscopy and energy-dispersive X-ray spectroscopy. The results indicate the range of species encountered at different temperatures showed the necessity for a time dependent sampling method. The investigated aerosols were in the nanoparticle range. A clear pattern was found for the use of fission product deposits from the Phébus FPT3 experiments. In oxidising atmosphere in the temperature range of 450K to 910K a species of 2nm sized spherical caesium-rhenium particles were observed that coagulated to spherical about 20nm sized particles. Other elements such as Mo and Te occurred at various temperature ranges and atmospheres. Also the vaporisation of caesium molybdate, as predicted in thermodynamic equilibrium calculations, was examined.

Phenomenological Analysis of Melt Progression in LWRS: Proposal of an In-Vessel Corium Progression Scenario in the Unit 1 of Fukushima Dai-ichi Nuclear Power Plant

In the field of severe accident, the description of corium progression events is mainly carried out using integral calculation codes. However, these tools are usually based on bounding assumptions because of the high complexity of phenomena. The limitations associated with bounding situations [1] (e.g., steady-state situations and instantaneous whole core relocation in the lower head) led CEA to develop an alternative approach to improve the phenomenological description of the melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core meltdown to the molten core–concrete interaction (MCCI). This phenomenological approach is based on the available data (including learnings from TMI-2) on physical models and knowledge about the corium behavior. It provides emerging trends and best-estimate intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three-dimensional (3D), all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the Unit 1 of the Fukushima Dai-ichi Nuclear Power Plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on the available results derived from codes such as the MELCOR calculation code [2]. The core degradation could then follow different ways: (1) Axial progression of the debris and the molten fuel through the lower support plate, or (2) lateral progression of the molten fuel through the shroud. On the basis of the Bali program results [3] and the TMI-2 accident observations [4], this work is focused on the consequences of a lateral melt progression (not excluding an axial progression through the support plate). Analysis of the events and the associated time sequence will be detailed. Besides, this analysis identifies some number of issues. Random calculations and statistical analysis of the results could be performed with calculation codes such as LEONAR–PROCOR codes [5]. This work was presented in the frame of the OECD/NEA/CSNI Benchmark Study of the Accident at the Fukushima Dai-ichi Nuclear Power Station (BSAF) project [6]. During the years of 2012 and 2014, the purpose of this project was both to study, by means of severe accident codes, the Fukushima accident in the three crippled units, until 6 days from the reactor shutdown, and to give information about, in particular, the location and composition of core debris.

Severe accident progression in the BWR lower plenum and the modes of vessel failure

Most of our knowledge base on the severe accident progression in the lower plenum of LWRs is based on the data obtained from the TMI-2 accident. It should be recognized that the lower plenum of a BWR is very different from that of a PWR. Unlike the PWR, the BWR plenum is full of control rod guide tubes (CRGTs) with their axial structural variations. These CRGTs are arranged in a cellular fashion with each CRGT supporting 4 rod bundles. There are also a large number of instrument guide tubes (IGTs), each generally placed in the middle of 4CRGTs. Both the CRGTs and IGTs traverse the thick vessel bottom wall and are welded to their extensions which come to bottom of the core. The core-melt progression in the lower plenum is controlled by the structures present and they, in turn, influence the timings and the modes of vessel failure for a BWR.The uranium oxide–zirconium oxide core melt formed in the 4 fuel bundles is directed by the structure below toward the water regions in-between the 4 CRGTs. The FCI will take place in those water regions and some particulate debris will be created, although there is insufficient water for quenching the melt. A FCI may occur inside a CRGT if and when the melt enters the CRGT at its top opening or the melt in the water region between the four CRGTs breaches the wall of the CRGT.The important issue is whether the welding holding the IGT inside the vessel will fail and the bottom part of the IGT falls out creating a hole in the vessel with release of water and melt/particulate debris from the vessel to the dry well of the BWR containment. Similarly, the failure of CRGT could have water and melt/particulate debris coming out of the vessel. These modes of vessel failure appear to be credible and they could occur before any large-scale melting and melt pool convection takes place. These modes of vessel failure and the melt release to the containment will have very different consequences than those generated by the other modes of vessel failure.Such BWR plenum melt progression scenarios have been considered in this paper. Some results of analyses performed at KTH have been described. We believe that the issues raised are important enough to consider a set of experiments for verification and validation of the melt progression in a BWR plenum. Such experiments are proposed.

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.

ASTEC validation on TMI-2 and LOFT LP-FP-2

The accident that occurred at the TMI-2 plant and the LP-FP-2 experiment conducted in the integral facility LOFT (Idaho National Laboratory) provided unique and valuable data for severe accident code assessment. Both the TMI-2 accident and the LP-FP-2 experiment were characterized by the loss of primary coolant with consequent core uncovery and heat-up that resulted in severe core damage. Both transients were terminated by core reflood with enhanced hydrogen generation and molten core material slumping into the lower plenum in case of TMI-2.These accidental transients have been analyzed with ASTEC V2 system code with the aim to validate and verify the coupling between the ICARE core degradation module and the CESAR module employed for the evaluation of the thermal-hydraulic behaviour of the primary circuit. The improved debris bed and magma models available in the latest version V2.0 of the code have been used to describe the late-phase of core melt progression. The comparison of code results with the available data has highlighted the globally good capability of the code to simulate the thermal-hydraulics of the primary circuit and the core melt progression during both early and late degradation phases, while code model weaknesses are observed in the analysis of the core reflood phase.

Development of CHF models for inner and outer RPV gaps in a meltdown severe accident

During a severe accident, the core melt relocates in the lower head and a hemispherical narrow gap may appear between the crust and the lower head because of the different material expansion ratio. The existence of this gap is very important to the integrity of the lower head. Based on the counter current flow limitation (CCFL) between the vapor phase and the liquid phase, a CHF model was developed to predict the CHF in hemispherical narrow gap. The CHF model developed was validated by the test data of Park and Köhler. The effect of key parameters, including the system pressure, radius of melt, and gap size, on the CHF were investigated. And the TMI-2 accident was also calculated by using the CHF formula. Moreover, based on the interface separation model, an analytical CHF model was developed to predict the CHF on the outer surface of the lower head. The predicted CHF was compared with the experimental data of ULPU-V. It indicated that the CHF models developed for the inner and outer CHF could predict the CHF well.

A THERP/ATHEANA Analysis of the Latent Operator Error in Leaving EFW Valves Closed in the TMI-2 Accident

This paper aims at performing a human reliability analysis using THERP (Technique for Human Error Prediction) and ATHEANA (Technique for Human Error Analysis) to develop a qualitative and quantitative analysis of the latent operator error in leaving EFW (emergency feed-water) valves closed in the TMI-2 accident. The accident analysis has revealed a series of unsafe actions that resulted in permanent loss of the unit. The integration between THERP and ATHEANA is developed in a way such as to allow a better understanding of the influence of operational context on human errors. This integration provides also, as a result, an intermediate method with the following features: (1) it allows the analysis of the action arising from the plant operational context upon the operator (as in ATHEANA), (2) it determines, as a consequence from the prior analysis, the aspects that most influence the context, and (3) it allows the change of these aspects into factors that adjust human error probabilities (as in THERP). This integration provides a more realistic and comprehensive modeling of accident sequences by considering preaccidental and postaccidental contexts, which, in turn, can contribute to more realistic PSA (Probabilistic Safety Assessment) evaluations and decision making.

Radionuclide Release to Stagnant Water in Fukushima-1 Nuclear Power Plant

After the severe accident in the Fukushima-1 nuclear power plant, a large amount of contaminated stagnant water has been produced in turbine buildings and surrounding areas. This rapid communication reports the calculation of the radionuclide inventory in the core, the collection of the measured inventory in the stagnant water, and the estimation of the radionuclide release ratios from the core to the stagnant water. The present evaluation is based on data obtained before June 3, 2011. It was revealed that the release ratios of tritium, iodine and cesium were several tens of percent, while those of strontium and barium were smaller by one or two orders of magnitude. These release ratios of the Fukushima accident were equivalent to those of the TMI-2 accident.

Looking to the future – an unanticipated event with potential new implications for packaging, transport and storage of radioactive material

PATRAM 2010 was convened in October 2010, with the theme ‘Looking to the future’. At that time, the emphasis on all of the papers was upon the then current situation worldwide in the peaceful application of radioactive materials. It included, as have previous PATRAM symposia, discussion of the development and application of regulations for the safe transport of radioactive material to and from ‘normal’ applications. Little if any focus was on the transport of materials arising from major accidents involving facilities using nuclear and other radioactive materials. In the past, the nuclear facilities devoted to the generation of electricity have experienced two significant events: unit 2 of Three Mile Island (TMI-2) in the USA and Chernobyl in the former Soviet Union. The TMI-2 accident resulted in significant clean-up of radioactive materials at the site, which involved processing of the contaminated and activated materials into forms that could be safely packaged, transported and stored or disposed. The Chernobyl accident resulted in much greater damage, and most of the contaminated and activated materials remain entombed on site. Since PATRAM 2010 occurred, the world has experienced the massive earthquake and tsunami, which significantly damaged the Fukushima Daiichi nuclear power plant (NPP).

Evaluation of recirculation sump performance for OPR1000 plant: Part I debris transport during the blow-down phase of LOCA

After TMI-2 accident, long-term core cooling management takes more importance rather than short-term management since probabilistic safety assessment performed revealed that long-term management had higher risk than the risk from short-term management. Regarding to this, since 1992, blockage of sump suction strainer was taken a place in Barseback Unit 2 (Sweden, BWR) and the United States Nuclear Regulatory Commission (USNRC) took into consideration as a General Safety Issue 191 (GSI-191), “PWR Sump Blockage,” in 1998 and opening an investigation. As a response to GSI-191 the Nuclear Energy Institute (NEI) submitted a recirculation sump evaluation methodology called NEI 04–07. In this methodology 0.75 was recommended for blow-down transport to lower containment based on the study on a boiling water reactor sump clogging issue and engineering judgment. USNRC quantitatively evaluated the blow-down transport in the safety evaluation report to NEI 04–07, and concluded the recommended blow-down transport fraction in NEI 04–07 was sufficiently conservative. However, the safety evaluation report to NEI 04–07 includes many values in evaluation steps which depend much on the containment configuration and engineering judgment. In particular the dependency on the plant type limits the generality of the USNRC’s conclusion on blow-down transport, when the considered plant is different from the volunteer plant. This study provides a modified evaluation method and results of debris blow-down transport for an Optimized Power Reactor 1000 MWe (OPR1000) in Korea. The modified method includes more realistic physical background with less uncertainty while providing consistent result with NEI and USNRC methods.

Development of linked analysis system for PWR severe accidents utilizing RELAP5 and MAAP4

The purpose of this study is to develop a severe accident (SA) analysis method that is more reliable thorough transferring the physical status of the plant predicted by RELAP5 computer code to MAAP4 computer code. The methodology of the linkage analysis is developed and the criterion of linkage time is suggested to utilize the RELAP5 thermal–hydraulic calculation to the maximum degree possible and thereby guarantee the continuity of calculation for hydrogen generation. The MAAP4 calculations after data transfer show the physically proper results based on RELAP5 data. Comparison with other code results for TMI-2 accident reveals that the result from the RELAP5–MAAP4 linked analysis lay in the span given by a number of results of TMI calculation from other SA code systems. The results of this study are expected to improve the SA analysis methodology by analyzing an SA scenario with more reliable thermal–hydraulic initial conditions.

Recent advances in ASTEC validation on circuit thermal-hydraulic and core degradation

Within the SARNET network of excellence in the 6th Framework Programme of the European Commission, the severe accident integral code ASTEC, jointly developed by IRSN (France) and GRS (Germany), has been validated against international experiments to evaluate the suitability and capability of new or improved models implemented in successive code versions up to V1.3rev2, delivered in December 2007. This paper focuses on the code applications concerning circuit thermal-hydraulics and core degradation to integral and separate-effect experiments: for the CESAR thermal-hydraulic module, BETHSY 9.1 b, PACTEL ISP 33 and T2.1, PMK2-SBLOCA, LOFT-LP-FP-2; for the DIVA core degradation module, CORA-13 and -W2, QUENCH-11 and -13, LOFT-LP-FP-2, Phébus FPT-4, FARO L14 and L28, LIVE-L1, OLHF-1, FOREVER EC2. Besides, the TMI-2 accident has been analyzed using the CESAR and DIVA modules in a coupling mode. The emphasis was put on the following new or improved models: i.e. in CESAR, reflooding of an intact core, condensation in the pressurizer, sub-critical break flow correlation, and new pressurizer spray model; in DIVA, corium behaviour in the lower head and lower head mechanical failure. For thermal-hydraulics in the circuits, good results have been obtained with ASTEC on the three integral experiments that cover various thermal-hydraulic flow regimes: LOFT-LP-FP-2 in Western PWR geometry and the two PACTEL experiments in VVER-440 geometry. These good results have been confirmed by the validation done on several BETHSY integral tests. For core degradation, the ASTEC results are good for early-phase models of core heat-up, oxidation and hydrogen production (before any quenching phase) on different CORA and QUENCH experiments and on LOFT-LP-FP-2. For the in-vessel late-phase, the results can be considered as good regarding debris bed melting (Phébus FPT-4), corium fragmentation at slump into vessel lower plenum (FARO), molten pool behaviour in lower plenum (LIVE-L1), and vessel lower head mechanics (OLHF-1 and FOREVER EC2). Furthermore, the first two phases of the TMI-2 accident before core reflooding are very well calculated by ASTEC. The main remaining modelling weaknesses concern the reflooding of a degraded core and the corresponding hydrogen production. The implementation of detailed magma 2D relocation models in the new series of ASTEC V2 versions (the first one being released mid-2009) will allow a more realistic simulation of late phase phenomena up to the failure of the lower head.

An assessment of effectiveness of core exit temperatures with respect to PWR core damage state using RELAP/SCDAPSIM/MOD3.4

Core exit temperature measurements play a very important role in the development and implementation of Severe Accident Management Guidelines. For example, the transition from Emergency Operating Procedures to the Severe Accident Management Guidelines may be triggered by core exit temperatures exceeding set limits. Core exit temperatures may also be used to determine if core cooling has been re-established when different management techniques have been applied. However, core exit temperatures may not always be reliable indicators of the state of the core. For that reason, a systematic study of the relationship between core exit temperatures and the state of the core for different accident conditions and different reactor designs has been started using the detailed system thermal hydraulic and severe accident analysis code package, RELAP/SCDAPSIM. In this paper, the results of the first part of the study is presented and discussed. The paper includes (a) a brief description of the RELAP/SCDAPSIM package, (b) insights from two full plant calculations, a Station Blackout transient for the Surry PWR and the TMI-2 accident analysis, and (c) the results from a series of sensitivity studies to relate core exit temperatures to core damage states for idealized core uncovery, heating, and reflooding scenarios for a typical PWR fuel assembly.

International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)

Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2) and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium) were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects. ISPs are providing unique material and benefits for some safety-related issues.