Fast Neutron Fluence Research Articles

Irradiation effects in tungsten-copper laminate composite

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410–780 °C and fast neutron fluences of 0.02–9.0 × 1025 n/m2, E > 0.1 MeV, 0.0039–1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22 °C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22 °C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile.

Influence of technological defects and neutron irradiation on the mechanical properties of single-crystal molybdenum pipes

The state of single-crystal molybdenum pipes irradiated in an IVV-2M reactor at the peak temperature of 1400°C for 1700 and 3200 h by the fast neutron fluence of 3 × 1020 and 5 × 1020 n cm–2, respectively, is investigated. The aim is to analyze the shape and surface state of pipes, the metallography of the structure, the short-time mechanical properties, and the character of destruction of the ring pipe specimen under testing. Radiation strengthening of molybdenum single-crystal is not registered. The strength parameters of the irradiated material are the level of initial values. The plastic properties of samples approach zero under room temperature tests. When the temperature is above 250°C, the total tensile strain of the sample exceeds 15% and ductile fracture with neck thinning to zero takes place. The perfect structure of an irradiated single crystal is preserved. Technological defects such as microcracks on the inner surface of pipes may act as stress concentrators.

Bayesian nonparametric analysis of crack growth rates in irradiated austenitic stainless steels in simulated BWR environments

To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (σYS-irr), stress intensity factor (K), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.

Quantification of dislocations densities in zirconium hydride by X-ray line profile analysis

Zirconium-based components in nuclear power plants are embrittled by precipitates of δ zirconium hydride, which involves a martensitic-type transformation of the hexagonal α-Zr lattice into the face-centered cubic Zr sublattice of the hydride. As a result, the hydride precipitates have a complex and heavily distorted internal structure that manifests as broad peaks in X-ray diffraction experiments. By a detailed analysis of the peak widths measured for different crystal planes we have found that most of this broadening is the result of dislocations. The analysis also showed that δ-hydride has very anisotropic mechanical elastic properties, in agreement with ab-initio simulations presented in the literature.Provided with this peak-broadening model, we have quantified dislocation densities within δ-hydrides precipitated in several Zr alloys, by analyzing previously published X-ray diffraction experiments performed at three synchrotron X-ray sources. The specimens investigated correspond to components affected by different hydride embrittling processes, namely: (i) samples from various components, charged in the laboratory with H contents in the ∼250 wt ppm range, (ii) laboratory-produced hydride blisters in Zr2.5%Nb pressure tubes; and (iii) Zircaloy-4 specimens machined from cooling channels of Atucha I nuclear power plant after 10 years in-service, containing ∼140 wt ppm of equivalent H content and subjected to an estimated fast neutron fluence of ∼1022 neutrons/cm2. Results show that dislocations densities in the δ-hydrides are large (5–20 × 1015 cm−2) and vary among the different specimens. We also found that dislocations densities in the hydride are proportional to the fraction of hydrides already formed in the matrix, which was interpreted as the effect of matrix hardness in the precipitate structure.

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 1025 n/m2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 1025 n/m2 (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 1025 n/m2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.

Reactor neutrons irradiation effect on the photocatalytic activity of SnO2 thin films

SnO2 thin films synthesized by sol-gel are irradiated with reactor neutrons up to fast neutron fluence of 9.6 × 1017 neutrons cm–2 at 40°C. The influence of defects generated by neutrons irradiation, through the properties modification, on the photo-catalytic activity of SnO2 films is investigated. It is found that the photoactivity of the irradiated films is enhanced after reactor neutrons irradiation. An improvement of 41% is observed for the sample irradiated at a neutron fluence of 9.6 × 1017 neutrons cm–2. This is attributed to several parameters modified by the reactor neutron irradiation principally the crystallite size and space charge region which are closely related to the photocatalytic performance.

Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak

We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.

An LEU-loaded long-life innovative sodium-cooled fast reactor (iSFR) with novel and passive safety devices

This paper presents the neutronics conceptual design of a long-life 392.6 MWth innovative sodium-cooled fast reactor (iSFR). The iSFR, which was derived based on Korean Prototype Generation-IV sodium cooled fast reactor (PGSFR), was intended as an alternative fast reactor option while waiting for the availability and readiness of the fuel recycling technology in the Republic of Korea. The iSFR core was designed for a lifetime of over 20 EFPYs (effective full power years) and an average core discharged burnup greater than 100 GWd/MTHM. It was loaded with annular metallic low enriched uranium (LEU) fuels, which were arranged in eight rings of hexagonal assemblies and surrounded by three rings of PbO radial reflectors. The core neutronic performance and characteristics such as core lifetime, power profiles and thermal–hydraulics analysis, kinetics parameters and reactivity feedback coefficients, and cladding fast neutron fluence were analyzed by using McCARD Monte Carlo code and MATRA-LMR thermal–hydraulics sub-channel code. A balance of reactivity (BOR) analysis was also investigated to preliminary assess the safety capability of the iSFR against several unprotected transient scenarios. In addition, two new unique passive safety devices named FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment) were also introduced and discussed in this paper.

Unconventional Electronic Properties of Mg2Si Thermoelectrics Revealed by Fast-Neutron-Irradiation Doping

In this work, we systematically investigated the electrical resistivity, Hall, and magnetoresistance effects of Al-doped Mg2Si thermoelectrics, irradiated with a fluence of fast neutrons and consequently isochronally annealed at moderate high temperatures up to 500 °C. We found that the fast-neutron bombardment itself only slightly modified the electronic properties. Meanwhile, a series of the postirradiation high-temperature anneals affected the properties dramatically. Thus, after annealing at temperatures of 275–325 °C, Mg2Si:Al showed pronounced jumps in both the electrical resistivity and the Hall constant values by several orders of magnitude. This unexpected electronic transition corresponded to a significant variation in the carrier concentration. We proposed that this unusual electronic transition may be related to temperature-assisted chemical bonding of the radiation defects that can involve free charge carriers in the sample, thereby tuning dramatically its transport properties. We proposed a ...

Core design optimization and analysis of the Purdue Novel Modular Reactor (NMR-50)

A BWR-based SMR called the Novel Modular Reactor (NMR-50) is being developed at Purdue University. NMR takes the advantages of the two-phase flow driving head, which allows a much smaller and simpler reactor pressure vessel (RPV) compared to the integral PWRs. In this study, through a systematic step-wise optimization approach including a simulated annealing based optimization method, an optimum core design that meets a 10-year cycle length with a minimum fuel cost while satisfying safety related criteria was derived and analyzed. The lattice code CASMO-4, the whole core analysis code PARCS and the thermal–hydraulics code RELAP5 were used to perform calculations from pin cell up to whole core depletion calculations. The NMR-50 optimized core design is able to achieve a 10.2year cycle length with an average fuel enrichment of 4.61wt% of 235U in a 10×10 lattice fuel assembly. The minimum critical power ratio (MCPR) and the maximum fuel linear power density (MFLPD) during the cycle are 1.99 and 18.25kW/m, respectively, providing large margins to thermal design constraints. The NMR-50 control system design is able to provide a sufficient cold shutdown margin of 1.7%. With its small reactor core size, large negative void coefficient, and low operating thermal neutron flux, an enhanced xenon stability characteristic is possible. Peak fast neutron fluence of 8.8×1021n/cm2 was below the industry standard limit, which from extensive plant data records, should not be a major concern to channel distortions from a radiation damage point of view.

Vacancy-related defects in n-type Si implanted with a rarefied microbeam of accelerated heavy ions in the MeV range

Deep level transient spectroscopy (DLTS) has been used to study vacancy-related defects formed in bulk n-type Czochralski-grown silicon after implantation of accelerated heavy ions: 6.5MeV O, 10.5MeV Si, 10.5MeV Ge, and 11MeV Er in the single ion regime with fluences from 109cm−2 to 1010cm−2 and a direct comparison made with defects formed in the same material irradiated with 0.7MeV fast neutron fluences up to 1012cm−2. A scanning ion microprobe was used as the ion implantation tool of n-Cz:Si samples prepared as Schottky diodes, while the ion beam induced current (IBIC) technique was utilized for direct ion counting. The single acceptor state of the divacancy V2(−/0) is the most prominent defect state observed in DLTS spectra of n-CZ:Si samples implanted by selected ions and the sample irradiated by neutrons. The complete suppression of the DLTS signal related to the double acceptor state of divacancy, V2(=/−) has been observed in all samples irradiated by ions and neutrons. Moreover, the DLTS peak associated with formation of the vacancy-oxygen complex VO in the neutron irradiated sample was also completely suppressed in DLTS spectra of samples implanted with the raster scanned ion microbeam. The reason for such behaviour is twofold, (i) the local depletion of the carrier concentration in the highly disordered regions, and (ii) the effect of the microprobe-assisted single ion implantation. The activation energy for electron emission for states assigned to the V2(−/0) defect formed in samples implanted by single ions follows the Meyer–Neldel rule. An increase of the activation energy is strongly correlated with increasing ion mass.

Estimation of the fast neutron fluence in Laguna Verde RPV-Steel specimens simulating their irradiation in a TRIGA reactor

The main purpose of this paper is to predict the fast neutron fluence value expected at four Laguna Verde Nuclear Power Plant RPV-steel specimens, when they are loaded in Mexican TRIGA Mark III reactor. Numerical simulation is done by neutronic calculation with MCNP5 Monte Carlo code for a detailed model of this research reactor core. This calculation will enable the estimation of the irradiation time needed for aging the specimens quickly. In the first part of this work, the MCNP5 developed model was validated through different operating conditions of the TRIGA reactor. The core was simulated to analyze the behavior of the keff when all the control rods are fully withdrawn and fully inserted in cold conditions (293 K); an analysis of the behavior of keff when only the regulating control rod is fully withdrawn is also included. Real operating conditions of the reactor at a power of 1 MWth in steady state were also simulated, in order to obtain the neutron flux in the six irradiation holes of the TRIGA reactor. The calculation of the axial neutron flux in each irradiation hole was validated applying the Neutron Activation Analysis technique (NAA), to calculate the activity of different materials that were irradiated in the TRIGA reactor; the results were compared with the activity calculated with MCNP5 for each material. Afterwards, the fast neutron flux that the RPV-steel specimens received in the experimental facility E−16 was also calculated with MCNP5. Finally, the irradiation time of the specimens in the TRIGA was estimated in order to obtain values of the fast neutron fluence at 24, 32 and 48 Effective Full Power Years (EFPY).

Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV orE>0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center) to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV) equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV) received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90). In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002). This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

Radiation annealing of radiation embrittlement of the reactor pressure vessel steel

Influence of neutron irradiation on RPV steel degradation are examined with reference to the possible reasons of the substantial experimental data scatter and furthermore - nonstandard (non-monotonous) and oscillatory embrittlement behavior. In our glance this phenomenon may be explained by presence of the wavelike component in the embrittlement kinetics. We suppose that the main factor affecting steel anomalous embrittlement is fast neutron intensity (dose rate or flux), flux effect manifestation depends on state-of-the-art fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. Data on radiation damage change including through the ex-service RPVs taking into account chemical factor, fast neutron fluence and neutron flux were obtained and analyzed. In our opinion controversy in the estimation on neutron flux on radiation degradation impact may be explained by presence of the wavelike component in the embrittlement kinetics. Therefore flux effect manifestation depends on fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. Moreover as a hypothesis we suppose that at some stages of irradiation damaged metal have to be partially restored by irradiation i.e. neutron bombardment. Nascent during irradiation structure undergo occurring once or periodically transformation in a direction both degradation and recovery of the initial properties. According to our hypothesis at some stage(s) of metal structure degradation neutron bombardment became recovering factor. As a result oscillation arise that in tern lead to enhanced data scatter.

Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

One of the major limiting factors to nuclear reactors lifetime is the radiation- induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I 2 S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I 2 S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I 2 S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center- peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I 2 S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

Self-Recovering Section of RPV Steel Radiation Embrittlement Kinetics as Indication of Material Smart Behavior

Influence of neutron irradiation on reactor pressure vessel (RPV) steel degradation are examined with reference to the possible reasons of the substantial experimental data scatter and furthermore -nonstandard (non-monotonous) and oscillatory embrittlement behavior. In our glance this phenomenon may be explained by presence of the wavelike recovering component in the embrittlement kinetics. We suppose that the main factor affecting steel anomalous embrittlement is fast neutron intensity (dose rate or flux), flux effect manifestation depends on state-of-the-art fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. Data on radiation damage change including through the ex-service RPVs taking into account chemical factor, fast neutron fluence and neutron flux were obtained and analyzed. In our opinion controversy in the estimation on neutron flux on radiation degradation impact may be explained by presence of the wavelike component in the embrittlement kinetics. Therefore flux effect manifestation depends on fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. As a result of dose rate effect manifestation peripheral RPV’s zones in some range of fluencies have to be damaged to a large extent than situated closely to core. Moreover as a hypothesis we suppose that at some stages of irradiation damaged metal have to be partially restored by irradiation i.e. neutron bombardment. Nascent during irradiation structure undergo occurring once or periodically transformation in a direction both degradation and recovery of the initial properties. According to our hypothesis at some stagem(s) of metal structure degradation neutron bombardment became recovering factor. Therefore self-recovering section of RPV steel radiation embrittlement kinetics is an indication of material smart behavior. As a result oscillation arise that intern lead to enhanced data scatter. Disclosure of the steel degradation oscillating is a sign of the steel structure cyclic self-recovery transformation as it take place in self-organization processes. This assumption has received support through the discovery of the similar “anomalous” data in scientific publications and by means of own additional experiments. Data obtained stimulate looking-for ways to management of the structural steel radiation stability (for example, by means of nano-structure modification for radiation defects annihilation intensification) for creation of the intelligent self-recovering material. Expected results: - radiation degradation theory and mechanisms development, - more adequate models of the radiation embrittlement elaboration, - methods and facility development for usage data of the accelerated materials irradiation for forecasting of their capacity for work in realistic (practical) circumstances of operation, - search of the ways for creating of the stable under neutron irradiation self-recovery smart materials.

Approaches to Accounting and Prediction of Fast Neutron Fluence on VVER Pressure Vessels for Estimation of RPV Residual Lifetime in Compliance with Russian Utility's Procedure

The Paper describes a new Russian Utility's regulatory document (RD EO) which has been recently developed and implemented since the beginning of 2013. This RD EO includes the procedure of RPV FNF monitoring and provides recommendations on how to predict fluence over the design lifetime taking into account results of FNF monitoring. The basic method of RPV neutron fluence monitoring is neutron transport calculations of FR in the vicinity of the RPV. Reliability of the calculation results should be validated by ex-vessel neutron-activation measurements, which were performed during different fuel cycles with different core loadings including new types of fuel.

Atomic structure of cascades of atomic displacements in metals and alloys after different types of radiation

Using the methods of field ion microscopy, we studied radiation induced defects on an atomically clean surface and within a subsurface volume of platinum initiated by the interaction of neutron (E > 0.1MeV) and Ar+ beams (E = 30 keV). It is shown that the interaction of fast neutrons (E > 0.1 MeV) F = 6.7-1021 m-2, F = 3.5-1022 m-2 with matter leads to the formation in the amount of platinum such as radiation damage which occur after ion irradiation by beams of charged Ar+ ions with E = 30 keV, F = 1020 ion/m2. They are observed at a depth of about 1.5-2 nm irradiated under the surface of Pt by ions Ar+. Thus, we have carried out modeling of neutron impact with matter when replacing the neutron beam by an ion beam that causes the same radiation damage in the bulk of the material. Experimental results on atomic-spatial investigation of radiative defect formation in surface layers of materials, initiated by neutron bombardment (of Pt, E > 0.1 MeV) and ion implantation (in Cu3Au: E = 40 keV, F = 1020 ion/m2, j = 10-3 A/cm2), are considered. Quantitative estimates obtained for the size, shape, and volume fraction of cascades of atomic displacements formed under various types of irradiation in the surface layers of the materials. It is showing that the average size of radiation clusters after irradiation of platinum to a fast neutron fluence of 6.7-1022 m-2 (E > 0.1 MeV) is about 3.2 nm. The experimentally established average size of a radiation cluster (disordered zone) in the alloy after ion bombardment is 4×4×1.5 nm.

Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.